On 29 July 2013, a shipment of nuclear material has left Italy and
was delivered to the United States of America.

The news is confirmed by influential members of the Italian
Government and in particular by the deputy minister of the Interior,
Filippo Bubbico.

We are journalists and we have documented the transport and write
in a local weekly magazine: "L'Independente Lucano" but,
for the occasion, we published an article in a newspaper prestigious
nationally: the weekly magazine "Oggi" of the publishing
group RCS (the first publishing group of newspapers in Italy).

Many people in Italy is concerned to know what happened to the
radioactive material taken from Basilicata and destined for the
United States.

We think it is also in the interest of the people of the United
States know where this material and what precautions have been taken
to its transport into U.S. territory.

Confident in the liberal and democratic tradition of the United
States, we then interviewed your agency "U.S. NRC" through
a series of emails and receives answer in which they confirm “no
records” related to nuclear material shipment from Italy.

U.S. NRC suggest to ask at NNSA and today we sent email to them..

Always known the respect and protection that the United States of
America reserves the free press, we can not explain why U.S. Agencies
had difficults to answer at our simple questions..

For this reason, Distinguished Mr. President Obama, we ask for
your intervention so that we can know when, where and by what
security guarantees, the radioactive cargo from Italy has passed in
the United States.

Dear Sirs and authority
delegated to the management of radioactive materials,

the day July 29, 2013 a shipment of radioactive material has
passed through the Italian streets.

Party Itrec center located in the town of Rotondella (Province of
Matera, Italy), it was escorted up to the military airport of Gioia
del Colle (province of Bari, Italy).

The Italian authorities have stated that it was about 1.2 kg of
uranium dioxide enriched to 91%. They specified that the final
destination of the cargo was the United States of America and that
the material was delivered to the final destination.

The transport, Italian authorities say was covered by state
secrecy. For this reason, local authorities and the population were
not warned.

We have carefully checked without being able to have confirmation.
In particular, we would like to know which site / city is today the
transported material and which path followed within the United
States.

Ask these questions because we know that the uranium dioxide is a
weakly radioactive material and we have verified that the U.S.
agencies: U.S. NRC (United States Nuclear Regulatory Commission), for
the "civil" and NNSA (National Nuclear Security
Administration) belonging to the DOE (Department of Energy), for the
"military" have never had difficulty in providing
information on these kinds of shipments.

In fact, on public sites of the two agencies, there are many
programs that affect communication, shipping and handling of enriched
uranium.

In truth, U.S. NRC has already replied to our question, confirming
that they do not is no transport of nuclear material from Italy.

We can not doubt the statements of the Italian Government,
therefore, will NNSA to know the details of this transport together
with EURATOM (EURATOM SUPPLY AGENCY) and IAEA (International Atomic
Energy Agency) which are international bodies that oversee and govern
the management and transport of radioactive material which is
strategic enriched uranium 235.

We believe that the Italian and U.S. citizens have a right to know
exactly what happened in the transport of 29 July last year and we
believe that the competent authorities have no difficulty in
providing all the elements necessary to reassure the populations
concerned.

We are confident that as the authorities will want to respond
quickly despite the holiday period, since these are highly sensitive
issues for the health and safety of citizens.

We greatly appreciate the policy of transparency always confirmed
by the United States of America and the grand opening always shown
towards the free press from that great country.

Subpart A--General Provisions

§ 71.0 Purpose and scope.

(1) Requirements for packaging, preparation for shipment, and
transportation of licensed material; and

(2) Procedures and standards for NRC approval of packaging and
shipping procedures for fissile material and for a quantity of other
licensed material in excess of a Type A quantity.

(b) The packaging and transport of licensed material are also
subject to other parts of this chapter (e.g., 10 CFR parts 20, 21,
30, 40, 70, and 73) and to the regulations of other agencies (e.g.,
the U.S. Department of Transportation (DOT) and the U.S. Postal
Service)1
having jurisdiction over means of transport. The requirements of this
part are in addition to, and not in substitution for, other
requirements.

(c) The regulations in this part apply to any licensee authorized
by specific or general license issued by the Commission to receive,
possess, use, or transfer licensed material, if the licensee delivers
that material to a carrier for transport, transports the material
outside the site of usage as specified in the NRC license, or
transports that material on public highways. No provision of this
part authorizes possession of licensed material.

(d)(1) Exemptions from the requirement for license in § 71.3 are
specified in § 71.14. General licenses for which no NRC package
approval is required are issued in §§ 71.20 through 71.23. The
general license in § 71.17 requires that an NRC certificate of
compliance or other package approval be issued for the package to be
used under this general license.

(2) Application for package approval must be completed in
accordance with subpart D of this part, demonstrating that the design
of the package to be used satisfies the package approval standards
contained in subpart E of this part, as related to the tests of
subpart F of this part.

(3) A licensee transporting licensed material, or delivering
licensed material to a carrier for transport, shall comply with the
operating control requirements of subpart G of this part; the quality
assurance requirements of subpart H of this part; and the general
provisions of subpart A of this part, including DOT regulations
referenced in § 71.5.

(e) The regulations of this part apply to any person holding, or
applying for, a certificate of compliance, issued pursuant to this
part, for a package intended for the transportation of radioactive
material, outside the confines of a licensee's facility or authorized
place of use.

(f) The regulations in this part apply to any person required to
obtain a certificate of compliance, or an approved compliance plan,
pursuant to part 76 of this chapter, if the person delivers
radioactive material to a common or contract carrier for transport or
transports the material outside the confines of the person's plant or
other authorized place of use.

(g) This part also gives notice to all persons who knowingly
provide to any licensee, certificate holder, quality assurance
program approval holder, applicant for a license, certificate, or
quality assurance program approval, or to a contractor, or
subcontractor of any of them, components, equipment, materials, or
other goods or services, that relate to a licensee's, certificate
holder's, quality assurance program approval holder's, or applicant's
activities subject to this part, that they may be individually
subject to NRC enforcement action for violation of § 71.8.

§ 71.1 Communications and records.

(a) Except where otherwise specified, all communications and
reports concerning the regulations in this part and applications
filed under them should be sent by mail addressed: ATTN: Document
Control Desk, Director, Division of Spent Fuel Storage and
Transportation, Office of Nuclear Material Safety and Safeguards,
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, by
hand delivery to the NRC's offices at 11555 Rockville Pike,
Rockville, Maryland; or, where practicable, by electronic submission,
for example, via Electronic Information Exchange, or CD-ROM.
Electronic submissions must be made in a manner that enables the NRC
to receive, read, authenticate, distribute, and archive the
submission, and process and retrieve it a single page at a time.
Detailed guidance on making electronic submissions can be obtained by
visiting the NRC’s Web site at
http://www.nrc.gov/site-help/e-submittals.html; by e-mail to
MSHD.Resource@nrc.gov; or by writing the Office of
Information Services, U.S. Nuclear Regulatory Commission, Washington,
DC 20555–0001. The guidance discusses, among other topics, the
formats the NRC can accept, the use of electronic signatures, and the
treatment of nonpublic information. If the submission date falls on a
Saturday, Sunday, or a Federal holiday, the next Federal working day
becomes the official due date.

(b) Each record required by this part must be legible throughout
the retention period specified by each Commission regulation. The
record may be the original or a reproduced copy or a microform
provided that the copy or microform is authenticated by authorized
personnel and that the microform is capable of producing a clear copy
throughout the required retention period. The record may also be
stored in electronic media with the capability for producing legible,
accurate, and complete records during the required retention period.
Records such as letters, drawings, and specifications must include
all pertinent information such as stamps, initials, and signatures.
The licensee shall maintain adequate safeguards against tampering
with and loss of records.

§ 71.2 Interpretations.

Except as specifically authorized by the Commission in writing, no
interpretation of the meaning of the regulations in this part by any
officer or employee of the Commission, other than a written
interpretation by the General Counsel, will be recognized to be
binding upon the Commission.

§ 71.3 Requirement for license.

Except as authorized in a general license or a specific license
issued by the Commission, or as exempted in this part, no licensee
may--

(a) Deliver licensed material to a carrier for transport; or

(b) Transport licensed material.

§ 71.4 Definitions.

The following terms are as defined here for the purpose of this
part. To ensure compatibility with international transportation
standards, all limits in this part are given in terms of dual units:
The International System of Units (SI) followed or preceded by U.S.
standard or customary units. The U.S. customary units are not exact
equivalents but are rounded to a convenient value, providing a
functionally equivalent unit. For the purpose of this part, either
unit may be used.

A1
means the maximum activity of special form radioactive material
permitted in a Type A package. This value is either listed in
Appendix A, Table A-1, of this part, or may be derived in accordance
with the procedures prescribed in Appendix A of this part.

A2
means the maximum activity of radioactive material, other than
special form material, LSA, and SCO material, permitted in a Type A
package. This value is either listed in Appendix A, Table A-1, of
this part, or may be derived in accordance with the procedures
prescribed in Appendix A of this part.

Carrier means a person engaged in the transportation of
passengers or property by land or water as a common, contract, or
private carrier, or by civil aircraft.

Certificate holder means a person who has been issued a
certificate of compliance or other package approval by the
Commission.

Certificate of Compliance (CoC) means the certificate
issued by the Commission under subpart D of this part which approves
the design of a package for the transportation of radioactive
material.

Close reflection by water means immediate contact by
water of sufficient thickness for maximum reflection of neutrons.

Consignment means each shipment of a package or groups of
packages or load of radioactive material offered by a shipper for
transport.

Containment system means the assembly of components of
the packaging intended to retain the radioactive material during
transport.

Conveyance means:

(1) For transport by public highway or rail any transport vehicle
or large freight container;

(2) For transport by water any vessel, or any hold, compartment,
or defined deck area of a vessel including any transport vehicle on
board the vessel; and

(3) For transport by any aircraft.

Criticality Safety Index (CSI) means the dimensionless
number (rounded up to the next tenth) assigned to and placed on the
label of a fissile material package, to designate the degree of
control of accumulation of packages containing fissile material
during transportation. Determination of the criticality safety index
is described in §§ 71.22, 71.23, and 71.59.

Deuterium means, for the purposes of §§ 71.15 and
71.22, deuterium and any deuterium compounds, including heavy water,
in which the ratio of deuterium atoms to hydrogen atoms exceeds
1:5000.

DOT means the U.S. Department of Transportation.

Exclusive use means the sole use by a single consignor of
a conveyance for which all initial, intermediate, and final loading
and unloading are carried out in accordance with the direction of the
consignor or consignee. The consignor and the carrier must ensure
that any loading or unloading is performed by personnel having
radiological training and resources appropriate for safe handling of
the consignment. The consignor must issue specific instructions, in
writing, for maintenance of exclusive use shipment controls, and
include them with the shipping paper information provided to the
carrier by the consignor.

Fissile material means the radionuclides uranium-233,
uranium-235, plutonium-239, and plutonium-241, or any combination of
these radionuclides. Fissile material means the fissile nuclides
themselves, not material containing fissile nuclides. Unirradiated
natural uranium and depleted uranium and natural uranium or depleted
uranium, that has been irradiated in thermal reactors only, are not
included in this definition. Certain exclusions from fissile material
controls are provided in §71.15.

Graphite means, for the purposes of §§ 71.15 and 71.22,
graphite with a boron equivalent content less than 5 parts per
million and density greater than 1.5 grams per cubic centimeter.

Indian tribe means an Indian or Alaska Native tribe,
band, nation, pueblo, village, or community that the Secretary of the
Interior acknowledges to exist as an Indian tribe pursuant to the
Federally Recognized Indian Tribe List Act of 1994, 25 U.S.C. 479a.

Licensed material means byproduct, source, or special
nuclear material received, possessed, used, or transferred under a
general or specific license issued by the Commission pursuant to the
regulations in this chapter.

Low Specific Activity (LSA) material means radioactive
material with limited specific activity which is nonfissile or is
excepted under §71.15, and which satisfies the descriptions and
limits set forth below. Shielding materials surrounding the LSA
material may not be considered in determining the estimated average
specific activity of the package contents. LSA material must be in
one of three groups:

(1) LSA—I.

(i) Uranium and thorium ores, concentrates of uranium and thorium
ores, and other ores containing naturally occurring radioactive
radionuclides which are not intended to be processed for the use of
these radionuclides;

(ii) Solid unirradiated natural uranium or depleted uranium or
natural thorium or their solid or liquid compounds or mixtures;

(iii) Radioactive material for which the A2
value is unlimited; or

(iv) Other radioactive material in which the activity is
distributed throughout and the estimated average specific activity
does not exceed 30 times the value for exempt material activity
concentration determined in accordance with Appendix A.

(2) LSA—II.

(i) Water with tritium concentration up to 0.8 TBq/liter (20.0
Ci/liter); or

(ii) Other material in which the activity is distributed
throughout and the average specific activity does not exceed 10
–4A2/g
for solids and gases, and 10–5A2/g
for liquids.

(i) The radioactive material is distributed throughout a solid or
a collection of solid objects, or is essentially uniformly
distributed in a solid compact binding agent (such as concrete,
bitumen, ceramic, etc.);

(ii) The radioactive material is relatively insoluble, or it is
intrinsically contained in a relatively insoluble material, so that
even under loss of packaging, the loss of radioactive material per
package by leaching, when placed in water for 7 days, would not
exceed 0.1 A2;
and

(iii) The estimated average specific activity of the solid does
not exceed 2 x 10–3A2/g.

Low toxicity alpha emitters means natural uranium,
depleted uranium, natural thorium; uranium-235, uranium-238,
thorium-232, thorium-228 or thorium-230 when contained in ores or
physical or chemical concentrates or tailings; or alpha emitters with
a half-life of less than 10 days.

Maximum normal operating pressure means the maximum gauge
pressure that would develop in the containment system in a period of
1 year under the heat condition specified in §71.71(c)(1), in the
absence of venting, external cooling by an ancillary system, or
operational controls during transport.

(2) Type A package means a Type A packaging together with its
radioactive contents. A Type A package is defined and must comply
with the DOT regulations in 49 CFR part 173.

(3) Type B package means a Type B packaging together with its
radioactive contents. On approval, a Type B package design is
designated by NRC as B(U) unless the package has a maximum normal
operating pressure of more than 700 kPa (100 lbs/in2)
gauge or a pressure relief device that would allow the release of
radioactive material to the environment under the tests specified in
§71.73 (hypothetical accident conditions), in which case it will
receive a designation B(M). B(U) refers to the need for unilateral
approval of international shipments; B(M) refers to the need for
multilateral approval of international shipments. There is no
distinction made in how packages with these designations may be used
in domestic transportation. To determine their distinction for
international transportation, see DOT regulations in 49 CFR Part 173.
A Type B package approved before September 6, 1983, was designated
only as Type B. Limitations on its use are specified in §71.19.

Packaging means the assembly of components necessary to
ensure compliance with the packaging requirements of this part. It
may consist of one or more receptacles, absorbent materials, spacing
structures, thermal insulation, radiation shielding, and devices for
cooling or absorbing mechanical shocks. The vehicle, tie-down system,
and auxiliary equipment may be designated as part of the packaging.

Special form radioactive material means radioactive
material that satisfies the following conditions:

(1) It is either a single solid piece or is contained in a sealed
capsule that can be opened only by destroying the capsule;

(2) The piece or capsule has at least one dimension not less than
5 mm (0.2 in); and

(3) It satisfies the requirements of §71.75. A special form
encapsulation designed in accordance with the requirements of §71.4
in effect on June 30, 1983 (see 10 CFR part 71, revised as of January
1, 1983), and constructed before July 1, 1985, and a special form
encapsulation designed in accordance with the requirements of §71.4
in effect on March 31, 1996 (see 10 CFR part 71, revised as of
January 1, 1983), and constructed before April 1, 1998, may continue
to be used. Any other special form encapsulation must meet the
specifications of this definition.

Specific activity of a radionuclide means the
radioactivity of the radionuclide per unit mass of that nuclide. The
specific activity of a material in which the radionuclide is
essentially uniformly distributed is the radioactivity per unit mass
of the material.

Spent nuclear fuel or Spent fuel means fuel that has been
withdrawn from a nuclear reactor following irradiation, has undergone
at least 1 year's decay since being used as a source of energy in a
power reactor, and has not been chemically separated into its
constituent elements by reprocessing. Spent fuel includes the special
nuclear material, byproduct material, source material, and other
radioactive materials associated with fuel assemblies.

State means a State of the United States, the District of
Columbia, the Commonwealth of Puerto Rico, the Virgin Islands, Guam,
American Samoa, and the Commonwealth of the Northern Mariana Islands.

Surface Contaminated Object (SCO) means a solid object
that is not itself classed as radioactive material, but which has
radioactive material distributed on any of its surfaces. SCO must be
in one of two groups with surface activity not exceeding the
following limits:

(1) SCO-I: A solid object on which:

(i) The nonfixed contamination on the accessible surface averaged
over 300 cm2
(or the area of the surface if less than 300 cm2)
does not exceed 4 Bq/cm2
(10–4
microcurie/cm2)
for beta and gamma and low toxicity alpha emitters, or 0.4 Bq/cm2
(10–5
microcurie/cm2)
for all other alpha emitters;

(ii) The fixed contamination on the accessible surface averaged
over 300 cm2
(or the area of the surface if less than 300 cm2)
does not exceed 4 x 104
Bq/cm2 (1.0
microcurie/cm2)
for beta and gamma and low toxicity alpha emitters, or 4 x 103
Bq/cm2 (0.1
microcurie/cm2)
for all other alpha emitters; and

(iii) The nonfixed contamination plus the fixed contamination on
the inaccessible surface averaged over 300 cm2
(or the area of the surface if less than 300 cm2)
does not exceed 4 x 104
Bq/cm2 (1
microcurie/cm2 )
for beta and gamma and low toxicity alpha emitters, or 4 x 103
Bq/cm2 (0.1
microcurie/cm2)
for all other alpha emitters.

(2) SCO-II: A solid object on which the limits for SCO-I are
exceeded and on which:

(i) The nonfixed contamination on the accessible surface averaged
over 300 cm2
(or the area of the surface if less than 300 cm2)
does not exceed 400 Bq/cm2
(10–2
microcurie/cm2)
for beta and gamma and low toxicity alpha emitters or 40 Bq/cm2
(10–3
microcurie/cm2)
for all other alpha emitters;

(ii) The fixed contamination on the accessible surface averaged
over 300 cm2
(or the area of the surface if less than 300 cm2)
does not exceed 8 x 105
Bq/cm2 (20 microcuries/cm2)
for beta and gamma and low toxicity alpha emitters, or 8 x 104
Bq/cm2 (2
microcuries/cm2)
for all other alpha emitters; and

(iii) The nonfixed contamination plus the fixed contamination on
the inaccessible surface averaged over 300 cm2
(or the area of the surface if less than 300 cm2)
does not exceed 8 x 105
Bq/cm2 (20
microcuries/cm2)
for beta and gamma and low toxicity alpha emitters, or 8 x 104
Bq/cm2 (2
microcuries/cm2)
for all other alpha emitters.

Transport index (TI) means the dimensionless number
(rounded up to the next tenth) placed on the label of a package, to
designate the degree of control to be exercised by the carrier during
transportation. The transport index is the number determined by
multiplying the maximum radiation level in millisievert (mSv) per
hour at 1 meter (3.3 ft) from the external surface of the package by
100 (equivalent to the maximum radiation level in millirem per hour
at 1 meter (3.3 ft)).

Tribal official means the highest ranking individual that
represents Tribal leadership, such as the Chief, President, or Tribal
Council leadership.

Type A quantity means a quantity of radioactive material,
the aggregate radioactivity of which does not exceed A1
for special form radioactive material, or A2,
for normal form radioactive material, where A1
and A 2 are
given in Table A-1 of this part, or may be determined by procedures
described in Appendix A of this part.

Type B quantity means a quantity of radioactive material
greater than a Type A quantity.

Unirradiated uranium means uranium containing not more
than 2 x 103 Bq
of plutonium per gram of uranium-235, not more than 9 x 106
Bq of fission products per gram of uranium-235, and not more than 5 x
10–3 g of
uranium-236 per gram of uranium-235.

§ 71.5 Transportation of licensed material.

(a) Each licensee who transports licensed material outside the
site of usage, as specified in the NRC license, or where transport is
on public highways, or who delivers licensed material to a carrier
for transport, shall comply with the applicable requirements of the
DOT regulations in 49 CFR parts 107, 171 through 180, and 390
through 397, appropriate to the mode of transport.

(1) The licensee shall particularly note DOT regulations in the
following areas:

(i) Packaging—49 CFR part 173: subparts A, B, and I.

(ii) Marking and labeling—49 CFR part 172: subpart D; and §§
172.400 through 172.407 and §§ 172.436 through 172.441 of subpart
E.

(2) The licensee shall also note DOT regulations pertaining to the
following modes of transportation:

(i) Rail—49 CFR part 174: subparts A through D and K.

(ii) Air—49 CFR part 175.

(iii) Vessel—49 CFR part 176: subparts A through F and M.

(iv) Public Highway—49 CFR part 177 and parts 390 through 397.

(b) If DOT regulations are not applicable to a shipment of
licensed material, the licensee shall conform to the standards and
requirements of the DOT specified in paragraph (a) of this section to
the same extent as if the shipment or transportation were subject to
DOT regulations. A request for modification, waiver, or exemption
from those requirements, and any notification referred to in those
requirements, must be filed with, or made to, the Director, Office of
Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001.

§ 71.6 Information collection requirements: OMB
approval.

(a) The Nuclear Regulatory Commission has submitted the
information collection requirements contained in this part to the
Office of Management and Budget (OMB) for approval as required by the
Paperwork Reduction Act (44 U.S.C. 3501 et seq.). The NRC may not
conduct or sponsor, and a person is not required to respond to, a
collection of information unless it displays a currently valid OMB
control number. OMB has approved the information collection
requirements contained in this part under control number 3150-0008.

§ 71.7 Completeness and accuracy of information.

(a) Information provided to the Commission by a licensee,
certificate holder, or an applicant for a license or CoC; or
information required by statute or by the Commission's regulations,
orders, license or CoC conditions, to be maintained by the licensee
or certificate holder, must be complete and accurate in all material
respects.

(b) Each licensee, certificate holder, or applicant for a license
or CoC must notify the Commission of information identified by the
licensee, certificate holder, or applicant for a license or CoC as
having, for the regulated activity, a significant implication for
public health and safety or common defense and security. A licensee,
certificate holder, or an applicant for a license or CoC violates
this paragraph only if the licensee, certificate holder, or applicant
for a license or CoC fails to notify the Commission of information
that the licensee, certificate holder, or applicant for a license or
CoC has identified as having a significant implication for public
health and safety or common defense and security. Notification must
be provided to the Administrator of the appropriate Regional Office
within 2 working days of identifying the information. This
requirement is not applicable to information which is already
required to be provided to the Commission by other reporting or
updating requirements.

§ 71.8 Deliberate misconduct.

(a) This section applies to any--

(1) Licensee;

(2) Certificate holder;

(3) Quality assurance program approval holder;

(4) Applicant for a license, certificate, or quality assurance
program approval;

(5) Contractor (including a supplier or consultant) or
subcontractor, to any person identified in paragraph (a)(4) of this
section; or

(6) Employees of any person identified in paragraphs (a)(1)
through (a)(5) of this section.

(b) A person identified in paragraph (a) of this section who
knowingly provides to any entity, listed in paragraphs (a)(1) through
(a)(5) of this section, any components, materials, or other goods or
services that relate to a licensee's, certificate holder's, quality
assurance program approval holder's, or applicant's activities
subject to this part may not:

(1) Engage in deliberate misconduct that causes or would have
caused, if not detected, a licensee, certificate holder, quality
assurance program approval holder, or any applicant to be in
violation of any rule, regulation, or order; or any term, condition
or limitation of any license, certificate, or approval issued by the
Commission; or

(2) Deliberately submit to the NRC, a licensee, a certificate
holder, quality assurance program approval holder, an applicant for a
license, certificate or quality assurance program approval, or a
licensee's, applicant's, certificate holder's, or quality assurance
program approval holder's contractor or subcontractor, information
that the person submitting the information knows to be incomplete or
inaccurate in some respect material to the NRC.

(c) A person who violates paragraph (b)(1) or (b)(2) of this
section may be subject to enforcement action in accordance with the
procedures in 10 CFR part 2, subpart B.

(d) For the purposes of paragraph (b)(1) of this section,
deliberate misconduct by a person means an intentional act or
omission that the person knows:

(1) Would cause a licensee, certificate holder, quality assurance
program approval holder, or applicant for a license, certificate, or
quality assurance program approval to be in violation of any rule,
regulation, or order; or any term, condition, or limitation of any
license or certificate issued by the Commission; or

(2) Constitutes a violation of a requirement, procedure,
instruction, contract, purchase order, or policy of a licensee,
certificate holder, quality assurance program approval holder,
applicant, or the contractor or subcontractor of any of them.

§ 71.9 Employee protection.

(a) Discrimination by a Commission licensee, certificate holder,
an applicant for a Commission license or a CoC, or a contractor or
subcontractor of any of these, against an employee for engaging in
certain protected activities, is prohibited. Discrimination includes
discharge and other actions that relate to compensation, terms,
conditions, or privileges of employment. The protected activities are
established in section 211 of the Energy Reorganization Act of 1974,
as amended, and in general are related to the administration or
enforcement of a requirement imposed under the Atomic Energy Act of
1954, as amended, or the Energy Reorganization Act of 1974, as
amended.

(1) The protected activities include, but are not limited to:

(i) Providing the Commission or his or her employer information
about alleged violations of either of the statutes named in paragraph
(a) of this section or possible violations of requirements imposed
under either of those statutes;

(ii) Refusing to engage in any practice made unlawful under either
of the statutes named in paragraph (a) of this section or under these
requirements if the employee has identified the alleged illegality to
the employer;

(iii) Requesting the Commission to institute action against his or
her employer for the administration or enforcement of these
requirements;

(iv) Testifying in any Commission proceeding, or before Congress,
or at any Federal or State proceeding regarding any provision (or
proposed provision) of either of the statutes named in paragraph (a)
of this section; and

(v) Assisting or participating in, or is about to assist or
participate in, these activities.

(2) These activities are protected even if no formal proceeding is
actually initiated as a result of the employee's assistance or
participation.

(3) This section has no application to any employee alleging
discrimination prohibited by this section who, acting without
direction from his or her employer (or the employer's agent),
deliberately causes a violation of any requirement of the Energy
Reorganization Act of 1974, as amended, or the Atomic Energy Act of
1954, as amended.

(b) Any employee who believes that he or she has been discharged
or otherwise discriminated against by any person for engaging in
protected activities specified in paragraph (a)(1) of this section
may seek a remedy for the discharge or discrimination through an
administrative proceeding in the Department of Labor. The
administrative proceeding must be initiated within 180 days after an
alleged violation occurs. The employee may do this by filing a
complaint alleging the violation with the Department of Labor,
Employment Standards Administration, Wage and Hour Division. The
Department of Labor may order reinstatement, back pay, and
compensatory damages.

(c) A violation of paragraph (a), (e), or (f) of this section by a
Commission licensee, certificate holder, applicant for a Commission
license or a CoC, or a contractor or subcontractor of any of these
may be grounds for:

(1) Denial, revocation, or suspension of the license or the CoC;

(2) Imposition of a civil penalty on the licensee, applicant, or a
contractor or subcontractor of the licensee or applicant; or

(3) Other enforcement action.

(d) Actions taken by an employer, or others, which adversely
affect an employee may be predicated upon nondiscriminatory grounds.
The prohibition applies when the adverse action occurs because the
employee has engaged in protected activities. An employee's
engagement in protected activities does not automatically render him
or her immune from discharge or discipline for legitimate reasons or
from adverse action dictated by nonprohibited considerations.

(e)(1) Each licensee, certificate holder, and applicant for a
license or CoC must prominently post the current revision of NRC Form
3, "Notice to Employees," referenced in §19.11(c) of this
chapter. This form must be posted at locations sufficient to permit
employees protected by this section to observe a copy on the way to
or from their place of work. The premises must be posted not later
than 30 days after an application is docketed and remain posted while
the application is pending before the Commission, during the term of
the license or CoC, and for 30 days following license or CoC
termination.

(2) Copies of NRC Form 3 may be obtained by writing to the
Regional Administrator of the appropriate U.S. Nuclear Regulatory
Commission Regional Office listed in Appendix D to part 20 of this
chapter or by calling the NRC Publishing Services Branch at
301-415-5877
begin_of_the_skype_highlighting GRATIS 301-415-5877 end_of_the_skype_highlighting.

(f) No agreement affecting the compensation, terms, conditions, or
privileges of employment, including an agreement to settle a
complaint filed by an employee with the Department of Labor pursuant
to section 211 of the Energy Reorganization Act of 1974, as amended,
may contain any provision which would prohibit, restrict, or
otherwise discourage an employee from participating in a protected
activity as defined in paragraph (a)(1) of this section including,
but not limited to, providing information to the NRC or to his or her
employer on potential violations or other matters within NRC's
regulatory responsibilities.

[72 FR 63975, Nov. 14, 2007]

§ 71.10 Public inspection of application.

Applications for approval of a package design under this part,
which are submitted to the Commission, may be made available for
public inspection, in accordance with provisions of parts 2 and 9 of
this chapter. This includes an application to amend or revise an
existing package design, any associated documents and drawings
submitted with the application, and any responses to NRC requests for
additional information.

§ 71.11 Protection of Safeguards Information

Each licensee, certificate holder, or applicant for a Certificate
of Compliance for a transportation package for transport of
irradiated reactor fuel, strategic special nuclear material, a
critical mass of special nuclear material, or byproduct material in
quantities determined by the Commission through order or regulation
to be significant to the public health and safety or the common
defense and security, shall protect Safeguards Information against
unauthorized disclosure in accordance with the requirements in §
73.21 and the requirements of § 73.22 or § 73.23 of this chapter,
as applicable.

[73 FR 63572, Oct. 24, 2008]

Subpart B--Exemptions

Source: 69 FR 3786, Jan. 26, 2004, unless otherwise noted.

§ 71.12 Specific exemptions.

On application of any interested person or on its own initiative,
the Commission may grant any exemption from the requirements of the
regulations in this part that it determines is authorized by law and
will not endanger life or property nor the common defense and
security.

§ 71.13 Exemption of physicians.

Any physician licensed by a State to dispense drugs in the
practice of medicine is exempt from § 71.5 with respect to transport
by the physician of licensed material for use in the practice of
medicine. However, any physician operating under this exemption must
be licensed under 10 CFR part 35 or the equivalent Agreement State
regulations.

§ 71.14 Exemption for low-level materials.

(a) A licensee is exempt from all the requirements of this part
with respect to shipment or carriage of the following low-level
materials:

(1) Natural material and ores containing naturally occurring
radionuclides that are not intended to be processed for use of these
radionuclides, provided the activity concentration of the material
does not exceed 10 times the values specified in Appendix A, Table
A-2, of this part.

(2) Materials for which the activity concentration is not greater
than the activity concentration values specified in Appendix A, Table
A-2 of this part, or for which the consignment activity is not
greater than the limit for an exempt consignment found in Appendix A,
Table A-2, of this part.

(b) A licensee is exempt from all the requirements of this part,
other than §§ 71.5 and 71.88, with respect to shipment or carriage
of the following packages, provided the packages do not contain any
fissile material, or the material is exempt from classification as
fissile material under § 71.15:

(1) A package that contains no more than a Type A quantity of
radioactive material;

(2) A package transported within the United States that contains
no more than 0.74 TBq (20 Ci) of special form plutonium-244; or

(i) That the LSA or SCO material has an external radiation dose of
less than or equal to 10 mSv/h (1 rem/h), at a distance of 3 m from
the unshielded material; or

(ii) That the package contains only LSA-I or SCO-I material.

§ 71.15 Exemption from classification as fissile
material.

Fissile material meeting the requirements of at least one of the
paragraphs (a) through (f) of this section are exempt from
classification as fissile material and from the fissile material
package standards of §§ 71.55 and 71.59, but are subject to all
other requirements of this part, except as noted.

(a) Individual package containing 2 grams or less fissile
material.

(b) Individual or bulk packaging containing 15 grams or less of
fissile material provided the package has at least 200 grams of solid
nonfissile material for every gram of fissile material. Lead,
beryllium, graphite, and hydrogenous material enriched in deuterium
may be present in the package but must not be included in determining
the required mass for solid nonfissile material.

(i) There is at least 2000 grams of solid nonfissile material for
every gram of fissile material, and

(ii) There is no more than 180 grams of fissile material
distributed within 360 kg of contiguous nonfissile material.

(2) Lead, beryllium, graphite, and hydrogenous material enriched
in deuterium may be present in the package but must not be included
in determining the required mass of solid nonfissile material.

(d) Uranium enriched in uranium-235 to a maximum of 1 percent by
weight, and with total plutonium and uranium-233 content of up to 1
percent of the mass of uranium-235, provided that the mass of any
beryllium, graphite, and hydrogenous material enriched in deuterium
constitutes less than 5 percent of the uranium mass.

(e) Liquid solutions of uranyl nitrate enriched in uranium-235 to
a maximum of 2 percent by mass, with a total plutonium and
uranium-233 content not exceeding 0.002 percent of the mass of
uranium, and with a minimum nitrogen to uranium atomic ratio (N/U) of
2. The material must be contained in at least a DOT Type A package.

(f) Packages containing, individually, a total plutonium mass of
not more than 1000 grams, of which not more than 20 percent by mass
may consist of plutonium-239, plutonium-241, or any combination of
these radionuclides.

§ 71.16 [Reserved]

Subpart C--General Licenses

Source: 69 FR 3792, Jan. 26, 2004, unless otherwise noted.

§ 71.17 General license: NRC-approved package.

(a) A general license is issued to any licensee of the Commission
to transport, or to deliver to a carrier for transport, licensed
material in a package for which a license, certificate of compliance
(CoC), or other approval has been issued by the NRC.

(b) This general license applies only to a licensee who has a
quality assurance program approved by the Commission as satisfying
the provisions of subpart H of this part.

(c) This general license applies only to a licensee who--

(1) Has a copy of the CoC, or other approval of the package, and
has the drawings and other documents referenced in the approval
relating to the use and maintenance of the packaging and to the
actions to be taken before shipment;

(2) Complies with the terms and conditions of the license,
certificate, or other approval, as applicable, and the applicable
requirements of subparts A, G, and H of this part; and

(3) Before the licensee's first use of the package, submits in
writing to: ATTN: Document Control Desk, Director, Division of Spent
Fuel Storage and Transportation, Office of Nuclear Material Safety
and Safeguards, using an appropriate method listed in § 71.1(a), the
licensee's name and license number and the package identification
number specified in the package approval.

(d) This general license applies only when the package approval
authorizes use of the package under this general license.

(e) For a Type B or fissile material package, the design of which
was approved by NRC before April 1, 1996, the general license is
subject to the additional restrictions of § 71.19.

[75 FR 73945, Nov. 30, 2010]

§ 71.18 [Reserved]

§ 71.19 Previously approved package.

(a) [Reserved]

(b) A Type B(U) package, a Type B(M) package, or a fissile
material package, previously approved by the NRC but without the
designation "- 85" in the identification number of the NRC
CoC, may be used under the general license of § 71.17 with the
following additional conditions:

(1) Fabrication of the package is satisfactorily completed by
April 1, 1999, as demonstrated by application of its model number in
accordance with § 71.85(c);

(2) A package used for a shipment to a location outside the United
States is subject to multilateral approval as defined in DOT
regulations at 49 CFR 173.403; and

(3) A serial number which uniquely identifies each packaging which
conforms to the approved design is assigned to and legibly and
durably marked on the outside of each packaging.

(c) A Type B(U) package, a Type B(M) package, or a fissile
material package previously approved by the NRC with the designation
"-85" in the identification number of the NRC CoC, may be
used under the general license of § 71.17 with the following
additional conditions:

(1) Fabrication of the package must be satisfactorily completed by
December 31, 2006, as demonstrated by application of its model number
in accordance with § 71.85(c); and

(2) After December 31, 2003, a package used for a shipment to a
location outside the United States is subject to multilateral
approval as defined in DOT regulations at 49 CFR 173.403.

(d) NRC will approve modifications to the design and authorized
contents of a Type B package, or a fissile material package,
previously approved by NRC, provided--

(1) The modifications of a Type B package are not significant with
respect to the design, operating characteristics, or safe performance
of the containment system, when the package is subjected to the tests
specified in §§ 71.71 and 71.73;

(2) The modifications of a fissile material package are not
significant, with respect to the prevention of criticality, when the
package is subjected to the tests specified in §§ 71.71 and 71.73;
and

(3) The modifications to the package satisfy the requirements of
this part.

(e) NRC will revise the package identification number to designate
previously approved package designs as B, BF, AF, B(U), B(M), B(U)F,
B(M)F, B(U)-85, B(U)F-85, B(M)-85, B(M)F-85, or AF-85 as appropriate,
and with the identification number suffix "-96" after
receipt of an application demonstrating that the design meets the
requirements of this part.

§ 71.20 General license: DOT specification
container.

(a) A general license is issued to any licensee of the Commission
to transport, or to deliver to a carrier for transport, licensed
material in a specification container for fissile material or for a
Type B quantity of radioactive material as specified in DOT
regulations at 49 CFR parts 173 and 178.

(b) This general license applies only to a licensee who has a
quality assurance program approved by the Commission as satisfying
the provisions of subpart H of this part.

(c) This general license applies only to a licensee who--

(1) Has a copy of the specification; and

(2) Complies with the terms and conditions of the specification
and the applicable requirements of subparts A, G, and H of this part.

(d) This general license is subject to the limitation that the
specification container may not be used for a shipment to a location
outside the United States, except by multilateral approval, as
defined in DOT regulations at 49 CFR 173.403.

(e) This section expires October 1, 2008.

§ 71.21 General license: Use of foreign approved
package.

(a) A general license is issued to any licensee of the Commission
to transport, or to deliver to a carrier for transport, licensed
material in a package, the design of which has been approved in a
foreign national competent authority certificate, that has been
revalidated by DOT as meeting the applicable requirements of 49 CFR
171.12.

(b) Except as otherwise provided in this section, the general
license applies only to a licensee who has a quality assurance
program approved by the Commission as satisfying the applicable
provisions of subpart H of this part.

(c) This general license applies only to shipments made to or from
locations outside the United States.

(d) This general license applies only to a licensee who--

(1) Has a copy of the applicable certificate, the revalidation,
and the drawings and other documents referenced in the certificate,
relating to the use and maintenance of the packaging and to the
actions to be taken before shipment; and

(2) Complies with the terms and conditions of the certificate and
revalidation, and with the applicable requirements of subparts A, G,
and H of this part. With respect to the quality assurance provisions
of subpart H of this part, the licensee is exempt from design,
construction, and fabrication considerations.

§ 71.22 General license: Fissile material.

(a) A general license is issued to any licensee of the Commission
to transport fissile material, or to deliver fissile material to a
carrier for transport, if the material is shipped in accordance with
this section. The fissile material need not be contained in a package
which meets the standards of subparts E and F of this part; however,
the material must be contained in a Type A package. The Type A
package must also meet the DOT requirements of 49 CFR 173.417(a).

(b) The general license applies only to a licensee who has a
quality assurance program approved by the Commission as satisfying
the provisions of subpart H of this part.

(c) The general license applies only when a package's contents:

(1) Contain no more than a Type A quantity of radioactive
material; and

(2) Contain less than 500 total grams of beryllium, graphite, or
hydrogenous material enriched in deuterium.

(d) The general license applies only to packages containing
fissile material that are labeled with a CSI which:

(1) Has been determined in accordance with paragraph (e) of this
section;

(2) Has a value less than or equal to 10; and

(3) For a shipment of multiple packages containing fissile
material, the sum of the CSIs must be less than or equal to 50 (for
shipment on a nonexclusive use conveyance) and less than or equal to
100 (for shipment on an exclusive use conveyance).

(e)(1) The value for the CSI must be greater than or equal to the
number calculated by the following equation:

(2) The calculated CSI must be rounded up to the first decimal
place;

(3) The values of X, Y, and Z used in the CSI equation must be
taken from Tables 71-1 or 71-2, as appropriate;

(4) If Table 71-2 is used to obtain the value of X, then the
values for the terms in the equation for uranium-233 and plutonium
must be assumed to be zero; and

(5) Table 71-1 values for X, Y, and Z must be used to determine
the CSI if:

(i) Uranium-233 is present in the package;

(ii) The mass of plutonium exceeds 1 percent of the mass of
uranium-235;

(iii) The uranium is of unknown uranium-235 enrichment or greater
than 24 weight percent enrichment; or

(iv) Substances having a moderating effectiveness (i.e., an
average hydrogen density greater than H2O)
(e.g., certain hydrocarbon oils or plastics) are present in any form,
except as polyethylene used for packing or wrapping.

Fissile material mass mixed with moderating substances having
an average hydrogen density less than or equal to H2O
(grams)

Fissile material mass mixed with moderating substances having
an average hydrogen density greater than H2Oa
(grams)

235U (X)

60

38

233U (Y)

43

27

239Pu or
241Pu (Z)

37

24

a
When mixtures of moderating substances are present, the lower mass
limits shall be used if more than 15 percent of the moderating
substance has an average hydrogen density greater than H2O.

Table 71-2. Mass Limits for General License
Packages Containing Uranium-235 of Known Enrichment per § 71.22(e)

Uranium enrichment in weight percent of 235U
not exceeding

Fissile material mass of 235U
(X) (grams)

24

60

20

63

15

67

11

72

10

76

9.5

78

9

81

8.5

82

8

85

7.5

88

7

90

6.5

93

6

97

5.5

102

5

108

4.5

114

4

120

3.5

132

3

150

2.5

180

2

246

1.5

408

1.35

480

1

1,020

0.92

1,800

[69 FR 3786, Jan. 26, 2004; 69 FR 58038, Sept. 29, 2004]

§ 71.23 General license: Plutonium-beryllium
special form material.

(a) A general license is issued to any licensee of the Commission
to transport fissile material in the form of plutonium-beryllium
(Pu-Be) special form sealed sources, or to deliver Pu-Be sealed
sources to a carrier for transport, if the material is shipped in
accordance with this section. This material need not be contained in
a package which meets the standards of subparts E and F of this part;
however, the material must be contained in a Type A package. The Type
A package must also meet the DOT requirements of 49 CFR 173.417(a).

(b) The general license applies only to a licensee who has a
quality assurance program approved by the Commission as satisfying
the provisions of subpart H of this part.

(c) The general license applies only when a package's contents:

(1) Contain no more than a Type A quantity of radioactive
material; and

(2) Contain less than 1000 g of plutonium, provided that:
plutonium-239, plutonium-241, or any combination of these
radionuclides, constitutes less than 240 g of the total quantity
of plutonium in the package.

(d) The general license applies only to packages labeled with a
CSI which:

(1) Has been determined in accordance with paragraph (e) of this
section;

(2) Has a value less than or equal to 100; and

(3) For a shipment of multiple packages containing Pu-Be sealed
sources, the sum of the CSIs must be less than or equal to 50 (for
shipment on a nonexclusive use conveyance) and less than or equal to
100 (for shipment on an exclusive use conveyance).

(e)(1) The value for the CSI must be greater than or equal to the
number calculated by the following equation:

(2) The calculated CSI must be rounded up to the first decimal
place.

§ 71.24 [Reserved]

§ 71.25 [Reserved]

Subpart D--Application for Package Approval

§ 71.31 Contents of application.

(a) An application for an approval under this part must include,
for each proposed packaging design, the following information:

(1) A package description as required by § 71.33;

(2) A package evaluation as required by § 71.35; and

(3) A quality assurance program description, as required by §
71.37, or a reference to a previously approved quality assurance
program.

(b) Except as provided in § 71.13, an application for
modification of a package design, whether for modification of the
packaging or authorized contents, must include sufficient information
to demonstrate that the proposed design satisfies the package
standards in effect at the time the application is filed.

(c) The applicant shall identify any established codes and
standards proposed for use in package design, fabrication, assembly,
testing, maintenance, and use. In the absence of any codes and
standards, the applicant shall describe and justify the basis and
rationale used to formulate the package quality assurance program.

§ 71.33 Package description.

The application must include a description of the proposed package
in sufficient detail to identify the package accurately and provide a
sufficient basis for evaluation of the package. The description must
include --

(4) Extent of reflection, the amount and identity of nonfissile
materials used as neutron absorbers or moderators, and the atomic
ratio of moderator to fissile constituents;

(5) Maximum normal operating pressure;

(6) Maximum weight;

(7) Maximum amount of decay heat; and

(8) Identification and volumes of any coolants.

§ 71.35 Package evaluation.

The application must include the following:

(a) A demonstration that the package satisfies the standards
specified in subparts E and F of this part;

(b) For a fissile material package, the allowable number of
packages that may be transported in the same vehicle in accordance
with § 71.59; and

(c) For a fissile material shipment, any proposed special controls
and precautions for transport, loading, unloading, and handling and
any proposed special controls in case of an accident or delay.

§ 71.37 Quality assurance.

(a) The applicant shall describe the quality assurance program
(see Subpart H of this part) for the design, fabrication, assembly,
testing, maintenance, repair, modification, and use of the proposed
package.

(b) The applicant shall identify any specific provisions of the
quality assurance program that are applicable to the particular
package design under consideration, including a description of the
leak testing procedures.

§ 71.38 Renewal of a certificate of compliance
or quality assurance program approval.

(a) Except as provided in paragraph (b) of this section, each
Certificate of Compliance or Quality Assurance Program Approval
expires at the end of the day, in the month and year stated in the
approval.

(b) In any case in which a person, not less than 30 days before
the expiration of an existing Certificate of Compliance or Quality
Assurance Program Approval issued pursuant to the part, has filed an
application in proper form for renewal of either of those approvals,
the existing Certificate of Compliance or Quality Assurance Program
Approval for which the renewal application was filed shall not be
deemed to have expired until final action on the application for
renewal has been taken by the Commission.

(c) In applying for renewal of an existing Certificate of
Compliance or Quality Assurance Program Approval, an applicant may be
required to submit a consolidated application that incorporates all
changes to its program that, are incorporated by reference in the
existing approval or certificate, into as few referenceable documents
as reasonably achievable.

§ 71.39 Requirement for additional information.

The Commission may at any time require additional information in
order to enable it to determine whether a license, certificate of
compliance, or other approval should be granted, renewed, denied,
modified, suspended, or revoked.

Subpart E--Package Approval Standards

§ 71.41 Demonstration of compliance.

(a) The effects on a package of the tests specified in § 71.71
("Normal conditions of transport"), and the tests specified
in § 71.73 ("Hypothetical accident conditions"), and §
71.61 ("Special requirements for Type B packages containing more
than 105 A2"),
must be evaluated by subjecting a specimen or scale model to a
specific test, or by another method of demonstration acceptable to
the Commission, as appropriate for the particular feature being
considered.

(b) Taking into account the type of vehicle, the method of
securing or attaching the package, and the controls to be exercised
by the shipper, the Commission may permit the shipment to be
evaluated together with the transporting vehicle.

(c) Environmental and test conditions different from those
specified in §§ 71.71 and 71.73 may be approved by the Commission
if the controls proposed to be exercised by the shipper are
demonstrated to be adequate to provide equivalent safety of the
shipment.

(d) Packages for which compliance with the other provisions of
these regulations is impracticable shall not be transported except
under special package authorization. Provided the applicant
demonstrates that compliance with the other provisions of the
regulations is impracticable and that the requisite standards of
safety established by these regulations have been demonstrated
through means alternative to the other provisions, a special package
authorization may be approved for one-time shipments. The applicant
shall demonstrate that the overall level of safety in transport for
these shipments is at least equivalent to that which would be
provided if all the applicable requirements had been met.

[60 FR 50264, Sept. 28, 1995 as amended at 69 FR 3794, Jan. 26,
2004]

§ 71.43 General standards for all packages.

(a) The smallest overall dimension of a package may not be less
than 10 cm (4 in).

(b) The outside of a package must incorporate a feature, such as a
seal, that is not readily breakable and that, while intact, would be
evidence that the package has not been opened by unauthorized
persons.

(c) Each package must include a containment system securely closed
by a positive fastening device that cannot be opened unintentionally
or by a pressure that may arise within the package.

(d) A package must be made of materials and construction that
assure that there will be no significant chemical, galvanic, or other
reaction among the packaging components, among package contents, or
between the packaging components and the package contents, including
possible reaction resulting from inleakage of water, to the maximum
credible extent. Account must be taken of the behavior of materials
under irradiation.

(e) A package valve or other device, the failure of which would
allow radioactive contents to escape, must be protected against
unauthorized operation and, except for a pressure relief device, must
be provided with an enclosure to retain any leakage.

(f) A package must be designed, constructed, and prepared for
shipment so that under the tests specified in § 71.71 ("Normal
conditions of transport") there would be no loss or dispersal of
radioactive contents, no significant increase in external surface
radiation levels, and no substantial reduction in the effectiveness
of the packaging.

(g) A package must be designed, constructed, and prepared for
transport so that in still air at 38°C (100°F) and in the shade, no
accessible surface of a package would have a temperature exceeding
50°C (122°F) in a nonexclusive use shipment, or 85°C (185°F) in
an exclusive use shipment.

(h) A package may not incorporate a feature intended to allow
continuous venting during transport.

§ 71.45 Lifting and tie-down standards for all
packages.

(a) Any lifting attachment that is a structural part of a package
must be designed with a minimum safety factor of three against
yielding when used to lift the package in the intended manner, and it
must be designed so that failure of any lifting device under
excessive load would not impair the ability of the package to meet
other requirements of this subpart. Any other structural part of the
package that could be used to lift the package must be capable of
being rendered inoperable for lifting the package during transport,
or must be designed with strength equivalent to that required for
lifting attachments.

(b) Tie-down devices:

(1) If there is a system of tie-down devices that is a structural
part of the package, the system must be capable of withstanding,
without generating stress in any material of the package in excess of
its yield strength, a static force applied to the center of gravity
of the package having a vertical component of 2 times the weight of
the package with its contents, a horizontal component along the
direction in which the vehicle travels of 10 times the weight of the
package with its contents, and a horizontal component in the
transverse direction of 5 times the weight of the package with its
contents.

(2) Any other structural part of the package that could be used to
tie down the package must be capable of being rendered inoperable for
tying down the package during transport, or must be designed with
strength equivalent to that required for tie-down devices.

(3) Each tie-down device that is a structural part of a package
must be designed so that failure of the device under excessive load
would not impair the ability of the package to meet other
requirements of this part.

§ 71.47 External radiation standards for all
packages.

(a) Except as provided in paragraph (b) of this section, each
package of radioactive materials offered for transportation must be
designed and prepared for shipment so that under conditions normally
incident to transportation the radiation level does not exceed 2
mSv/h (200 mrem/h) at any point on the external surface of the
package, and the transport index does not exceed 10.

(b) A package that exceeds the radiation level limits specified in
paragraph (a) of this section must be transported by exclusive use
shipment only, and the radiation levels for such shipment must not
exceed the following during transportation:

(1) 2 mSv/h (200 mrem/h) on the external surface of the package,
unless the following conditions are met, in which case the limit is
10 mSv/h (1000 mrem/h):

(i) The shipment is made in a closed transport vehicle;

(ii) The package is secured within the vehicle so that its
position remains fixed during transportation; and

(iii) There are no loading or unloading operations between the
beginning and end of the transportation;

(2) 2 mSv/h (200 mrem/h) at any point on the outer surface of the
vehicle, including the top and underside of the vehicle; or in the
case of a flat-bed style vehicle, at any point on the vertical planes
projected from the outer edges of the vehicle, on the upper surface
of the load or enclosure, if used, and on the lower external surface
of the vehicle; and

(3) 0.1 mSv/h (10 mrem/h) at any point 2 meters (80 in) from the
outer lateral surfaces of the vehicle (excluding the top and
underside of the vehicle); or in the case of a flat-bed style
vehicle, at any point 2 meters (6.6 feet) from the vertical planes
projected by the outer edges of the vehicle (excluding the top and
underside of the vehicle); and

(4) 0.02 mSv/h (2 mrem/h) in any normally occupied space, except
that this provision does not apply to private carriers, if exposed
personnel under their control wear radiation dosimetry devices in
conformance with 10 CFR 20.1502.

(c) For shipments made under the provisions of paragraph (b) of
this section, the shipper shall provide specific written instructions
to the carrier for maintenance of the exclusive use shipment
controls. The instructions must be included with the shipping paper
information.

(d) The written instructions required for exclusive use shipments
must be sufficient so that, when followed, they will cause the
carrier to avoid actions that will unnecessarily delay delivery or
unnecessarily result in increased radiation levels or radiation
exposures to transport workers or members of the general public.

§ 71.51 Additional requirements for Type B
packages.

(a) A Type B package, in addition to satisfying the requirements
of §§ 71.41 through 71.47, must be designed, constructed, and
prepared for shipment so that under the tests specified in:

(1) Section 71.71 ("Normal conditions of transport"),
there would be no loss or dispersal of radioactive contents--as
demonstrated to a sensitivity of 10-6
A2 per hour, no
significant increase in external surface radiation levels, and no
substantial reduction in the effectiveness of the packaging; and

(2) Section 71.73 ("Hypothetical accident conditions"),
there would be no escape of krypton-85 exceeding 10 A2
in 1 week, no escape of other radioactive material exceeding a total
amount A2 in 1
week, and no external radiation dose rate exceeding 10 mSv/h (1
rem/h) at 1 m (40 in) from the external surface of the package.

(b) Where mixtures of different radionuclides are present, the
provisions of appendix A, paragraph IV of this part shall apply,
except that for Krypton-85, an effective A2
value equal to 10 A2
may be used.

(c) Compliance with the permitted activity release limits of
paragraph (a) of this section may not depend on filters or on a
mechanical cooling system.

(d) For packages which contain radioactive contents with activity
greater than 105
A2, the
requirements of § 71.61 must be met.

[60 FR 50264, Sept. 28, 1995 as amended at 69 FR 3794, Jan. 26,
2004]

§ 71.53 [Reserved]

[62 FR 5913, Feb. 10, 1997; 69 FR 3794, January 26, 2004]

§ 71.55 General requirements for fissile
material packages.

(a) A package used for the shipment of fissile material must be
designed and constructed in accordance with §§ 71.41 through 71.47.
When required by the total amount of radioactive material, a package
used for the shipment of fissile material must also be designed and
constructed in accordance with § 71.51.

(b) Except as provided in paragraph (c) or (g) of this section, a
package used for the shipment of fissile material must be so designed
and constructed and its contents so limited that it would be
subcritical if water were to leak into the containment system, or
liquid contents were to leak out of the containment system so that,
under the following conditions, maximum reactivity of the fissile
material would be attained:

(1) The most reactive credible configuration consistent with the
chemical and physical form of the material;

(2) Moderation by water to the most reactive credible extent; and

(3) Close full reflection of the containment system by water on
all sides, or such greater reflection of the containment system as
may additionally be provided by the surrounding material of the
packaging.

(c) The Commission may approve exceptions to the requirements of
paragraph (b) of this section if the package incorporates special
design features that ensure that no single packaging error would
permit leakage, and if appropriate measures are taken before each
shipment to ensure that the containment system does not leak.

(d) A package used for the shipment of fissile material must be so
designed and constructed and its contents so limited that under the
tests specified in § 71.71 ("Normal conditions of transport")
--

(1) The contents would be subcritical;

(2) The geometric form of the package contents would not be
substantially altered;

(3) There would be no leakage of water into the containment system
unless, in the evaluation of undamaged packages under § 71.59(a)(1),
it has been assumed that moderation is present to such an extent as
to cause maximum reactivity consistent with the chemical and physical
form of the material; and

(4) There will be no substantial reduction in the effectiveness of
the packaging, including:

(i) No more than 5 percent reduction in the total effective volume
of the packaging on which nuclear safety is assessed;

(ii) No more than 5 percent reduction in the effective spacing
between the fissile contents and the outer surface of the packaging;
and

(iii) No occurrence of an aperture in the outer surface of the
packaging large enough to permit the entry of a 10 cm (4 in) cube.

(e) A package used for the shipment of fissile material must be so
designed and constructed and its contents so limited that under the
tests specified in § 71.73 ("Hypothetical accident
conditions"), the package would be subcritical. For this
determination, it must be assumed that:

(1) The fissile material is in the most reactive credible
configuration consistent with the damaged condition of the package
and the chemical and physical form of the contents;

(2) Water moderation occurs to the most reactive credible extent
consistent with the damaged condition of the package and the chemical
and physical form of the contents; and

(3) There is full reflection by water on all sides, as close as is
consistent with the damaged condition of the package.

(f) For fissile material package designs to be transported by air:

(1) The package must be designed and constructed, and its contents
limited so that it would be subcritical, assuming reflection by 20 cm
(7.9 in) of water but no water inleakage, when subjected to
sequential application of:

(i) The free drop test in § 71.73(c)(1);

(ii) The crush test in § 71.73(c)(2);

(iii) A puncture test, for packages of 250 kg or more, consisting
of a free drop of the specimen through a distance of 3 m (120 in) in
a position for which maximum damage is expected at the conclusion of
the test sequence, onto the upper end of a solid, vertical,
cylindrical, mild steel probe mounted on an essentially unyielding,
horizontal surface. The probe must be 20 cm (7.9 in) in diameter,
with the striking end forming the frustum of a right circular cone
with the dimensions of 30 cm height, 2.5 cm top diameter, and a top
edge rounded to a radius of not more than 6 mm (0.25 in). For
packages less than 250 kg, the puncture test must be the same, except
that a 250 kg probe must be dropped onto the specimen which must be
placed on the surface; and

(iv) The thermal test in § 71.73(c)(4), except that the duration
of the test must be 60 minutes.

(2) The package must be designed and constructed, and its contents
limited, so that it would be subcritical, assuming reflection by 20
cm (7.9 in) of water but no water inleakage, when subjected to an
impact on an unyielding surface at a velocity of 90 m/s normal to the
surface, at such orientation so as to result in maximum damage. A
separate, undamaged specimen can be used for this evaluation.

(3) Allowance may not be made for the special design features in
paragraph (c) of this section, unless water leakage into or out of
void spaces is prevented following application of the tests in
paragraphs (f)(1) and (f)(2) of this section, and subsequent
application of the immersion test in § 71.73(c)(5).

(g) Packages containing uranium hexafluoride only are excepted
from the requirements of paragraph (b) of this section provided that:

(1) Following the tests specified in § 71.73 ("Hypothetical
accident conditions"), there is no physical contact between the
valve body and any other component of the packaging, other than at
its original point of attachment, and the valve remains leak tight;

(2) There is an adequate quality control in the manufacture,
maintenance, and repair of packagings;

(3) Each package is tested to demonstrate closure before each
shipment; and

(4) The uranium is enriched to not more than 5 weight percent
uranium-235.

§ 71.57 [Reserved]

§ 71.59 Standards for arrays of fissile material
packages.

(a) A fissile material package must be controlled by either the
shipper or the carrier during transport to assure that an array of
such packages remains subcritical. To enable this control, the
designer of a fissile material package shall derive a number "N"
based on all the following conditions being satisfied, assuming
packages are stacked together in any arrangement and with close full
reflection on all sides of the stack by water:

(1) Five times "N" undamaged packages with nothing
between the packages would be subcritical;

(2) Two times "N" damaged packages, if each package were
subjected to the tests specified in § 71.73 ("Hypothetical
accident conditions") would be subcritical with optimum
interspersed hydrogenous moderation; and

(3) The value of "N" cannot be less than 0.5.

(b) The CSI must be determined by dividing the number 50 by the
value of "N" derived using the procedures specified in
paragraph (a) of this section. The value of the CSI may be zero
provided that an unlimited number of packages are subcritical, such
that the value of "N" is effectively equal to infinity
under the procedures specified in paragraph (a) of this section. Any
CSI greater than zero must be rounded up to the first decimal place.

(c) For a fissile material package which is assigned a CSI value--

(1) Less than or equal to 50, that package may be shipped by a
carrier in a nonexclusive use conveyance, provided the sum of the
CSIs is limited to less than or equal to 50.

(2) Less than or equal to 50, that package may be shipped by a
carrier in an exclusive use conveyance, provided the sum of the CSIs
is limited to less than or equal to 100.

(3) Greater than 50, that package must be shipped by a carrier in
an exclusive use conveyance, provided the sum of the CSIs is limited
to less than or equal to 100.

[69 FR 3795, Jan. 26, 2004]

§ 71.61 Special requirements for Type B packages
containing more than 105A2.

A Type B package containing more than 105A2
must be designed so that its undamaged containment system can
withstand an external water pressure of 2 MPa (290 psi) for a period
of not less than 1 hour without collapse, buckling, or inleakage of
water.

[69 FR 3795, Jan. 26, 2004]

§ 71.63 Special requirement for plutonium
shipments.

Shipments containing plutonium must be made with the contents in
solid form, if the contents contain greater than 0.74 TBq (20 Ci) of
plutonium.

[69 FR 3795, Jan. 26, 2004]

§ 71.64 Special requirements for plutonium air
shipments.

(a) A package for the shipment of plutonium by air subject to §
71.88(a)(4), in addition to satisfying the requirements of §§ 71.41
through 71.63, as applicable, must be designed, constructed, and
prepared for shipment so that under the tests specified in --

(i) The containment vessel would not be ruptured in its
post-tested condition, and the package must provide a sufficient
degree of containment to restrict accumulated loss of plutonium
contents to not more than an A2
quantity in a period of 1 week;

(ii) The external radiation level would not exceed 10 mSv/h (1
rem/h) at a distance of 1 m (40 in) from the surface of the package
in its post-tested condition in air; and

(iii) A single package and an array of packages are demonstrated
to be subcritical in accordance with this part, except that the
damaged condition of the package must be considered to be that which
results from the plutonium accident tests in § 71.74, rather than
the hypothetical accident tests in § 71.73; and

(2) Section 71.74(c), there would be no detectable leakage of
water into the containment vessel of the package.

(b) With respect to the package requirements of paragraph (a),
there must be a demonstration or analytical assessment showing that
--

(1) The results of the physical testing for package qualification
would not be adversely affected to a significant extent by --

(i) The presence, during the tests, of the actual contents that
will be transported in the package; and

(ii) Ambient water temperatures ranging from 0.6°C (+33°F) to
38°C (+100°F) for those qualification tests involving water, and
ambient atmospheric temperatures ranging from -40°C (-40°F) to
+54°C (+130°F) for the other qualification tests.

(2) The ability of the package to meet the acceptance standards
prescribed for the accident condition sequential tests would not be
adversely affected if one or more tests in the sequence were deleted.

§ 71.65 Additional requirements.

The Commission may, by rule, regulation, or order, impose
requirements on any licensee, in addition to those established in
this part, as it deems necessary or appropriate to protect public
health or to minimize danger to life or property.

§ 71.71 Normal conditions of transport.

(a) Evaluation. Evaluation of each package design under
normal conditions of transport must include a determination of the
effect on that design of the conditions and tests specified in this
section. Separate specimens may be used for the free drop test, the
compression test, and the penetration test, if each specimen is
subjected to the water spray test before being subjected to any of
the other tests.

(b) Initial conditions. With respect to the initial
conditions for the tests in this section, the demonstration of
compliance with the requirements of this part must be based on the
ambient temperature preceding and following the tests remaining
constant at that value between -29°C (-20°F) and +38°C (+100°F)
which is most unfavorable for the feature under consideration. The
initial internal pressure within the containment system must be
considered to be the maximum normal operating pressure, unless a
lower internal pressure consistent with the ambient temperature
considered to precede and follow the tests is more unfavorable.

(c) Conditions and tests.

(1) Heat. An ambient temperature of 38°C (100°F) in
still air, and insolation according to the following table:

INSOLATION DATA

Form and location of surface

Total insolation for a 12-hour period (g cal/cm2)

Flat surfaces transported horizontally;

Base

None

Other surfaces

800

Flat surfaces not transported horizontally

200

Curved surfaces

400

(2) Cold. An ambient temperature
of -40°C (-40°F) in still air and shade.

(6) Water spray. A water spray that simulates exposure to
rainfall of approximately 5 cm/h (2 in/h) for at least 1 hour.

(7) Free drop. Between 1.5 and 2.5 hours after the
conclusion of the water spray test, a free drop through the distance
specified below onto a flat, essentially unyielding, horizontal
surface, striking the surface in a position for which maximum damage
is expected.

Criteria for Free Drop Test (Weight/Distance)

Package weight

Free drop distance

Kilograms

(Pounds)

Meters

(Feet)

Less than 5,000

(Less than 11,000)

1.2

(4)

5,000 to 10,000

(11,000 to 22,000)

0.9

(3)

10,000 to 15,000

(22,000 to 33,100)

0.6

(2)

More than 15,000

(More than 33,100)

0.3

(1)

(8) Corner drop. A free drop onto each corner of the
package in succession, or in the case of a cylindrical package onto
each quarter of each rim, from a height of 0.3 m (1 ft) onto a flat,
essentially unyielding, horizontal surface. This test applies only to
fiberboard, wood, or fissile material rectangular packages not
exceeding 50 kg (110 lbs) and fiberboard, wood, or fissile material
cylindrical packages not exceeding 100 kg (220 lbs).

(9) Compression. For packages weighing up to 5000 kg
(11,000 lbs), the package must be subjected, for a period of 24
hours, to a compressive load applied uniformly to the top and bottom
of the package in the position in which the package would normally be
transported. The compressive load must be the greater of the
following:

(i) The equivalent of 5 times the weight of the package; or

(ii) The equivalent of 13 kPa (2 lbf/in2)
multiplied by the vertically projected area of the package.

(10) Penetration. Impact of the hemispherical end of a
vertical steel cylinder of 3.2 cm (1.25 in) diameter and 6 kg (13
lbs) mass, dropped from a height of 1 m (40 in) onto the exposed
surface of the package that is expected to be most vulnerable to
puncture. The long axis of the cylinder must be perpendicular to the
package surface.

2
The package standards related to the tests in this subpart are
contained in subpart E of this part.

§ 71.73 Hypothetical accident conditions.

(a) Test procedures. Evaluation for hypothetical accident
conditions is to be based on sequential application of the tests
specified in this section, in the order indicated, to determine their
cumulative effect on a package or array of packages. An undamaged
specimen may be used for the water immersion tests specified in
paragraph (c)(6) of this section.

(b) Test conditions. With respect to the initial
conditions for the tests, except for the water immersion tests, to
demonstrate compliance with the requirements of this part during
testing, the ambient air temperature before and after the tests must
remain constant at that value between -29°C (-20°F) and +38°C
(+100°F) which is most unfavorable for the feature under
consideration. The initial internal pressure within the containment
system must be the maximum normal operating pressure, unless a lower
internal pressure, consistent with the ambient temperature assumed to
precede and follow the tests, is more unfavorable.

(c) Tests. Tests for hypothetical accident conditions
must be conducted as follows:

(1) Free Drop. A free drop of the specimen through a
distance of 9 m (30 ft) onto a flat, essentially unyielding,
horizontal surface, striking the surface in a position for which
maximum damage is expected.

(2) Crush. Subjection of the specimen to a dynamic crush
test by positioning the specimen on a flat, essentially unyielding
horizontal surface so as to suffer maximum damage by the drop of a
500-kg (1100-lb) mass from 9 m (30 ft) onto the specimen. The mass
must consist of a solid mild steel plate 1 m (40 in) by 1 m (40 in)
and must fall in a horizontal attitude. The crush test is required
only when the specimen has a mass not greater than 500 kg (1100 lb),
an overall density not greater than 1000 kg/m3
(62.4 lb/ft3)
based on external dimension, and radioactive contents greater than
1000 A2 not as
special form radioactive material. For packages containing fissile
material, the radioactive contents greater than 1000 A2
criterion does not apply.

(3) Puncture. A free drop of the specimen through a
distance of 1 m (40 in) in a position for which maximum damage is
expected, onto the upper end of a solid, vertical, cylindrical, mild
steel bar mounted on an essentially unyielding, horizontal surface.
The bar must be 15 cm (6 in) in diameter, with the top horizontal and
its edge rounded to a radius of not more than 6 mm (0.25 in), and of
a length as to cause maximum damage to the package, but not less than
20 cm (8 in) long. The long axis of the bar must be vertical.

(4) Thermal. Exposure of the specimen fully engulfed,
except for a simple support system, in a hydrocarbon fuel/air fire of
sufficient extent, and in sufficiently quiescent ambient conditions,
to provide an average emissivity coefficient of at least 0.9, with an
average flame temperature of at least 800°C (1475°F) for a period
of 30 minutes, or any other thermal test that provides the equivalent
total heat input to the package and which provides a time averaged
environmental temperature of 800°C. The fuel source must extend
horizontally at least 1 m (40 in), but may not extend more than 3 m
(10 ft), beyond any external surface of the specimen, and the
specimen must be positioned 1 m (40 in) above the surface of the fuel
source. For purposes of calculation, the surface absorptivity
coefficient must be either that value which the package may be
expected to possess if exposed to the fire specified or 0.8,
whichever is greater; and the convective coefficient must be that
value which may be demonstrated to exist if the package were exposed
to the fire specified. Artificial cooling may not be applied after
cessation of external heat input, and any combustion of materials of
construction, must be allowed to proceed until it terminates
naturally.

(5) Immersion--fissile material. For fissile material
subject to § 71.55, in those cases where water inleakage has not
been assumed for criticality analysis, immersion under a head of
water of at least 0.9 m (3 ft) in the attitude for which maximum
leakage is expected.

(6) Immersion--all packages. A separate, undamaged
specimen must be subjected to water pressure equivalent to immersion
under a head of water of at least 15 m (50 ft). For test purposes, an
external pressure of water of 150 kPa (21.7 lbf/in2)
gauge is considered to meet these conditions.

[69 FR 3795, Jan. 26, 2004]

§ 71.74 Accident conditions for air transport of
plutonium.

(a) Test conditions--Sequence of tests. A package must be
physically tested to the following conditions in the order indicated
to determine their cumulative effect.

(1) Impact at a velocity of not less than 129 m/sec (422 ft/sec)
at a right angle onto a flat, essentially unyielding, horizontal
surface, in the orientation (e.g., side, end, corner) expected to
result in maximum damage at the conclusion of the test sequence.

(2) A static compressive load of 31,800 kg (70,000 lbs) applied in
the orientation expected to result in maximum damage at the
conclusion of the test sequence. The force on the package must be
developed between a flat steel surface and a 5 cm (2 in) wide,
straight, solid, steel bar. The length of the bar must be at least as
long as the diameter of the package, and the longitudinal axis of the
bar must be parallel to the plane of the flat surface. The load must
be applied to the bar in a manner that prevents any members or
devices used to support the bar from contacting the package.

(3) Packages weighing less than 227 kg (500 lbs) must be placed on
a flat, essentially unyielding, horizontal surface, and subjected to
a weight of 227 kg (500 lbs) falling from a height of 3 m (10 ft) and
striking in the position expected to result in maximum damage at the
conclusion of the test sequence. The end of the weight contacting the
package must be a solid probe made of mild steel. The probe must be
the shape of the frustum of a right circular cone, 30 cm (12 in)
long, 20 cm (8 in) in diameter at the base, and 2.5 cm (1 in) in
diameter at the end. The longitudinal axis of the probe must be
perpendicular to the horizontal surface. For packages weighing 227 kg
(500 lbs) or more, the base of the probe must be placed on a flat,
essentially unyielding horizontal surface, and the package dropped
from a height of 3 m (10 ft) onto the probe, striking in the position
expected to result in maximum damage at the conclusion of the test
sequence.

(4) The package must be firmly restrained and supported such that
its longitudinal axis is inclined approximately 45° to the
horizontal. The area of the package that made first contact with the
impact surface in paragraph (a)(1) of this section must be in the
lowermost position. The package must be struck at approximately the
center of its vertical projection by the end of a structural steel
angle section falling from a height of at least 46 m (150 ft). The
angle section must be at least 1.8 m (6 ft) in length with equal legs
at least 13 cm (5 in) long and 1.3 cm (0.5 in) thick. The angle
section must be guided in such a way as to fall end-on, without
tumbling. The package must be rotated approximately 90° about its
longitudinal axis and struck by the steel angle section falling as
before.

(5) The package must be exposed to luminous flames from a pool
fire of JP-4 or JP-5 aviation fuel for a period of at least 60
minutes. The luminous flames must extend an average of at least 0.9 m
(3 ft) and no more than 3 m (10 ft) beyond the package in all
horizontal directions. The position and orientation of the package in
relation to the fuel must be that which is expected to result in
maximum damage at the conclusion of the test sequence. An alternate
method of thermal testing may be substituted for this fire test,
provided that the alternate test is not of shorter duration and would
not result in a lower heating rate to the package. At the conclusion
of the thermal test, the package must be allowed to cool naturally or
must be cooled by water sprinkling, whichever is expected to result
in maximum damage at the conclusion of the test sequence.

(6) Immersion under at least 0.9 m (3 ft) of water.

(b) Individual free-fall impact test.

(1) An undamaged package must be physically subjected to an impact
at a velocity not less than the calculated terminal free-fall
velocity, at mean sea level, at a right angle onto a flat,
essentially unyielding, horizontal surface, in the orientation (e.g.,
side, end, corner) expected to result in maximum damage.

(2) This test is not required if the calculated terminal free-fall
velocity of the package is less than 129 m/sec (422 ft/sec), or if a
velocity not less than either 129 m/sec (422 ft/sec) or the
calculated terminal free-fall velocity of the package is used in the
sequential test of paragraph (a)(1) of this section.

(c) Individual deep submersion test. An undamaged package must be
physically submerged and physically subjected to an external water
pressure of at least 4 MPa (600 lbs/in2).

§ 71.75 Qualification of special form
radioactive material.

(a) Special form radioactive materials must meet the test
requirements of paragraph (b) of this section. Each solid radioactive
material or capsule specimen to be tested must be manufactured or
fabricated so that it is representative of the actual solid material
or capsule that will be transported, with the proposed radioactive
content duplicated as closely as practicable. Any differences between
the material to be transported and the test material, such as the use
of non-radioactive contents, must be taken into account in
determining whether the test requirements have been met. In addition:

(1) A different specimen may be used for each of the tests;

(2) The specimen may not break or shatter when subjected to the
impact, percussion, or bending tests;

(3) The specimen may not melt or disperse when subjected to the
heat test;

(4) After each test, leaktightness or indispersibility of the
specimen must be determined by a method no less sensitive than the
leaching assessment procedure prescribed in paragraph (c) of this
section. For a capsule resistant to corrosion by water, and which has
an internal void volume greater than 0.1 milliliter, an alternative
to the leaching assessment is a demonstration of leaktightness of
x10-4
torr-liter/s (1.3xx10-4
atm-cm3/s)
based on air at 25°C (77°F) and one atmosphere differential
pressure for solid radioactive content, or x10-6
torr-liter/s (1.30xx10-6
atm-cm3/s) for
liquid or gaseous radioactive content; and

(5) A specimen that comprises or simulates radioactive material
contained in a sealed capsule need not be subjected to the
leaktightness procedure specified in this section, provided it is
alternatively subjected to any of the tests prescribed in
ISO/TR4826-1979(E), "Sealed radioactive sources leak test
methods" which is available from the American National Standards
Institute, 1430 Broadway, New York, N.Y. 10018.

(b) Test methods. (1) Impact Test. The specimen
must fall onto the target from a height of 9 m (30 ft) or greater in
the orientation expected to result in maximum damage. The target must
be a flat, horizontal surface of such mass and rigidity that any
increase in its resistance to displacement or deformation, on impact
by the specimen, would not significantly increase the damage to the
specimen.

(2) Percussion Test. (i) The specimen must be placed on a
sheet of lead that is supported by a smooth solid surface, and struck
by the flat face of a steel billet so as to produce an impact
equivalent to that resulting from a free drop of 1.4 kg (3 lbs)
through 1 m (40 in);

(ii) The flat face of the billet must be 25 millimeters (mm) (1
inch) in diameter with the edges rounded off to a radius of 3 mm±0.3
mm(.12 in±0.012 in);

(iii) The lead must be hardness number 3.5 to 4.5 on the Vickers
scale and thickness 25 mm (1 in) or greater, and must cover an area
greater than that covered by the specimen;

(iv) A fresh surface of lead must be used for each impact; and

(v) The billet must strike the specimen so as to cause maximum
damage.

(3) Bending test. (i) This test applies only to long,
slender sources with a length of 10 cm (4 inches) or greater and a
length to width ratio of 10 or greater;

(ii) The specimen must be rigidly clamped in a horizontal position
so that one half of its length protrudes from the face of the clamp;

(iii) The orientation of the specimen must be such that the
specimen will suffer maximum damage when its free end is struck by
the flat face of a steel billet;

(iv) The billet must strike the specimen so as to produce an
impact equivalent to that resulting from a free vertical drop of 1.4
kg (3 lbs) through 1 m (40 in); and

(v) The flat face of the billet must be 25 mm (1 inch) in diameter
with the edges rounded off to a radius of 3 mm±0.3 mm (.12 in±0.012
in).

(4) Heat test. The specimen must be heated in air to a
temperature of not less than 800°C (1475°F), held at that
temperature for a period of 10 minutes, and then allowed to cool.

(i) The specimen must be immersed for 7 days in water at ambient
temperature. The water must have a pH of 6-8 and a maximum
conductivity of 10 micromho per centimeter at 20° (68°F);

(ii) The water with specimen must then be heated to a temperature
of 50°C±5°C (122°F±9°F) and maintained at this temperature for
4 hours.

(iii) The activity of the water must then be determined;

(iv) The specimen must then be stored for at least 7 days in still
air of relative humidity not less than 90 percent at 30°C (86°F);

(v) The specimen must then be immersed in water under the same
conditions as in paragraph (c)(1)(i) of this section, and the water
with specimen must be heated to 50°C±5°C (122°F±9°F) and
maintained at that temperature for 4 hours;

(vi) The activity of the water must then be determined. The sum of
the activities determined here and in paragraph (c)(1)(iii) of this
section must not exceed 2 kilobecquerels (kBq) (0.05 microcurie
(µCi)).

(2) For encapsulated material —

(i) The specimen must be immersed in water at ambient temperature.
The water must have a pH of 6-8 and a maximum conductivity of 10
micromho per centimeter;

(ii) The water and specimen must be heated to a temperature of
50°C±5°C (122°F±9°F) and maintained at this temperature for 4
hours;

(iii) The activity of the water must then be determined;

(iv) The specimen must then be stored for at least 7 days in still
air at a temperature of 30°C (86°F) or greater;

(v) The process in paragraph (c)(2)(i), (ii), and (iii) of this
section must be repeated; and

(vi) The activity of the water must then be determined. The sum of
the activities determined here and in paragraph (c)(2)(iii) of this
section must not exceed 2 kilobecquerels (kBq) (0.05 microcurie
(Ci)).

(d) A specimen that comprises or simulates radioactive material
contained in a sealed capsule need not be subjected to —

(1) The impact test and the percussion test of this section,
provided that the specimen is alternatively subjected to the Class 4
impact test prescribed in ISO 2919-1980(e), "Sealed Radioactive
Sources Classification" (see § 71.75(a)(5) for statement of
availability); and

(2) The heat test of this section, provided the specimen is
alternatively subjected to the Class 6 temperature test specified in
the International Organization for Standardization document ISO
2919-1980(e), "Sealed Radioactive Sources Classification."

§ 71.77 Qualification of LSA-III Material

(a) LSA-III material must meet the test requirements of paragraph
(b) of this section. Any differences between the specimen to be
tested and the material to be transported must be taken into account
in determining whether the test requirements have been met.

(b) Leaching Test. (1) The specimen, representing no less
than the entire contents of the package, must be immersed for 7 days
in water at ambient temperature;

(2) The volume of water to be used in the test must be sufficient
to ensure that at the end of the test period the free volume of the
unabsorbed and unreacted water remaining will be at least 10% of the
volume of the specimen itself;

(3) The water must have an initial pH of 6-8 and a maximum
conductivity 10 micromho/cm at 20°C (68°F); and

(4) The total activity of the free volume of water must be
measured following the 7 day immersion test and must not exceed 0.1
A2.

Subpart G--Operating Controls and Procedures

§ 71.81 Applicability of operating controls and
procedures.

A licensee subject to this part, who, under a general or specific
license, transports licensed material or delivers licensed material
to a carrier for transport, shall comply with the requirements of
this subpart G, with the quality assurance requirements of subpart H
of this part, and with the general provisions of subpart A of this
part.

§ 71.83 Assumptions as to unknown properties.

When the isotopic abundance, mass, concentration, degree of
irradiation, degree of moderation, or other pertinent property of
fissile material in any package is not known, the licensee shall
package the fissile material as if the unknown properties have
credible values that will cause the maximum neutron multiplication.

§ 71.85 Preliminary determinations.

Before the first use of any packaging for the shipment of licensed
material --

(a) The licensee shall ascertain that there are no cracks,
pinholes, uncontrolled voids, or other defects that could
significantly reduce the effectiveness of the packaging;

(b) Where the maximum normal operating pressure will exceed 35 kPa
(5 lbf/in2)
gauge, the licensee shall test the containment system at an internal
pressure at least 50 percent higher than the maximum normal operating
pressure, to verify the capability of that system to maintain its
structural integrity at that pressure; and

(c) The licensee shall conspicuously and durably mark the
packaging with its model number, serial number, gross weight, and a
package identification number assigned by NRC. Before applying the
model number, the licensee shall determine that the packaging has
been fabricated in accordance with the design approved by the
Commission.

§ 71.87 Routine determinations.

Before each shipment of licensed material, the licensee shall
ensure that the package with its contents satisfies the applicable
requirements of this part and of the license. The licensee shall
determine that --

(a) The package is proper for the contents to be shipped;

(b) The package is in unimpaired physical condition except for
superficial defects such as marks or dents;

(c) Each closure device of the packaging, including any required
gasket, is properly installed and secured and free of defects;

(d) Any system for containing liquid is adequately sealed and has
adequate space or other specified provision for expansion of the
liquid;

(e) Any pressure relief device is operable and set in accordance
with written procedures;

(f) The package has been loaded and closed in accordance with
written procedures;

(g) For fissile material, any moderator or neutron absorber, if
required, is present and in proper condition;

(h) Any structural part of the package that could be used to lift
or tie down the package during transport is rendered inoperable for
that purpose, unless it satisfies the design requirements of §
71.45;

(i) The level of non-fixed (removable) radioactive contamination
on the external surfaces of each package offered for shipment is as
low as reasonably achievable, and within the limits specified in DOT
regulations in 49 CFR 173.443;

(j) External radiation levels around the package and around the
vehicle, if applicable, will not exceed the limits specified in §
71.47 at any time during transportation; and

(k) Accessible package surface temperatures will not exceed the
limits specified in § 71.43(g) at any time during transportation.

§ 71.88 Air transport of plutonium.

(a) Notwithstanding the provisions of any general licenses and
notwithstanding any exemptions stated directly in this part or
included indirectly by citation of 49 CFR chapter I, as may be
applicable, the licensee shall assure that plutonium in any form,
whether for import, export, or domestic shipment, is not transported
by air or delivered to a carrier for air transport unless:

(1) The plutonium is contained in a medical device designed for
individual human application; or

(2) The plutonium is contained in a material in which the specific
activity is less than or equal to the activity concentration values
for plutonium specified in Appendix A, Table A-2, of this part, and
in which the radioactivity is essentially uniformly distributed; or

(3) The plutonium is shipped in a single package containing no
more than an A2
quantity of plutonium in any isotope or form, and is shipped in
accordance with § 71.5; or

(4) The plutonium is shipped in a package specifically authorized
for the shipment of plutonium by air in the Certificate of Compliance
for that package issued by the Commission.

(b) Nothing in paragraph (a) of this section is to be interpreted
as removing or diminishing the requirements of § 73.24 of this
chapter.

(c) For a shipment of plutonium by air which is subject to
paragraph (a)(4) of this section, the licensee shall, through special
arrangement with the carrier, require compliance with 49 CFR 175.704,
U.S. Department of Transportation regulations applicable to the air
transport of plutonium.

[69 FR 3795, Jan. 26, 2004]

§ 71.89 Opening instructions.

Before delivery of a package to a carrier for transport, the
licensee shall ensure that any special instructions needed to safely
open the package have been sent to, or otherwise made available to,
the consignee for the consignee's use in accordance with 10 CFR
20.1906(e).

§ 71.91 Records.

(a) Each licensee shall maintain, for a period of 3 years after
shipment, a record of each shipment of licensed material not exempt
under § 71.10, showing where applicable --

(1) Identification of the packaging by model number and serial
number;

(2) Verification that there are no significant defects in the
packaging, as shipped;

(3) Volume and identification of coolant;

(4) Type and quantity of licensed material in each package, and
the total quantity of each shipment;

(5) For each item of irradiated fissile material --

(i) Identification by model number and serial number;

(ii) Irradiation and decay history to the extent appropriate to
demonstrate that its nuclear and thermal characteristics comply with
license conditions; and

(iii) Any abnormal or unusual condition relevant to radiation
safety;

(6) Date of the shipment;

(7) For fissile packages and for Type B packages, any special
controls exercised;

(8) Name and address of the transferee;

(9) Address to which the shipment was made; and

(10) Results of the determinations required by § 71.87 and by the
conditions of the package approval.

(b) Each certificate holder shall maintain, for a period of 3
years after the life of the packaging to which they apply, records
identifying the packaging by model number, serial number, and date of
manufacture.

(c) The licensee, certificate holder, and an applicant for a CoC,
shall make available to the Commission for inspection, upon
reasonable notice, all records required by this part. Records are
only valid if stamped, initialed, or signed and dated by authorized
personnel, or otherwise authenticated.

(d) The licensee, certificate holder, and an applicant for a CoC
shall maintain sufficient written records to furnish evidence of the
quality of packaging. The records to be maintained include results of
the determinations required by § 71.85; design, fabrication, and
assembly records; results of reviews, inspections, tests, and audits;
results of monitoring work performance and materials analyses; and
results of maintenance, modification, and repair activities.
Inspection, test, and audit records must identify the inspector or
data recorder, the type of observation, the results, the
acceptability, and the action taken in connection with any
deficiencies noted. These records must be retained for 3 years after
the life of the packaging to which they apply.

[69 FR 3796, Jan. 26, 2004]

§ 71.93 Inspection and tests.

(a) The licensee, certificate holder, and applicant for a CoC
shall permit the Commission, at all reasonable times, to inspect the
licensed material, packaging, premises, and facilities in which the
licensed material or packaging is used, provided, constructed,
fabricated, tested, stored, or shipped.

(b) The licensee, certificate holder, and applicant for a CoC
shall perform, and permit the Commission to perform, any tests the
Commission deems necessary or appropriate for the administration of
the regulations in this chapter.

(c) The certificate holder and applicant for a CoC shall notify
the NRC, in accordance with § 71.1, 45 days in advance of starting
fabrication of the first packaging under a CoC. This paragraph
applies to any packaging used for the shipment of licensed material
which has either--

§ 71.95 Reports.

(a) The licensee, after requesting the certificate holder's input,
shall submit a written report to the Commission of--

(1) Instances in which there is a significant reduction in the
effectiveness of any NRC-approved Type B or Type AF packaging during
use; or

(2) Details of any defects with safety significance in any
NRC-approved Type B or fissile material packaging, after first use.

(3) Instances in which the conditions of approval in the
Certificate of Compliance were not observed in making a shipment.

(b) The licensee shall submit a written report to the Commission
of instances in which the conditions in the certificate of compliance
were not followed during a shipment.

(c) Each licensee shall submit, in accordance with § 71.1, a
written report required by paragraph (a) or (b) of this section
within 60 days of the event or discovery of the event. The licensee
shall also provide a copy of each report submitted to the NRC to the
applicable certificate holder. Written reports prepared under other
regulations may be submitted to fulfill this requirement if the
reports contain all the necessary information, and the appropriate
distribution is made. Using an appropriate method listed in §
71.1(a), the licensee shall report to: ATTN: Document Control Desk,
Director, Division of Spent Fuel Storage and Transportation, Office
of Nuclear Material Safety and Safeguards. These written reports must
include the following:

(1) A brief abstract describing the major occurrences during the
event, including all component or system failures that contributed to
the event and significant corrective action taken or planned to
prevent recurrence.

(2) A clear, specific, narrative description of the event that
occurred so that knowledgeable readers conversant with the
requirements of part 71, but not familiar with the design of the
packaging, can understand the complete event. The narrative
description must include the following specific information as
appropriate for the particular event.

(i) Status of components or systems that were inoperable at the
start of the event and that contributed to the event;

(ii) Dates and approximate times of occurrences;

(iii) The cause of each component or system failure or personnel
error, if known;

(iv) The failure mode, mechanism, and effect of each failed
component, if known;

(v) A list of systems or secondary functions that were also
affected for failures of components with multiple functions;

(vi) The method of discovery of each component or system failure
or procedural error;

(vii) For each human performance-related root cause, a discussion
of the cause(s) and circumstances;

(viii) The manufacturer and model number (or other identification)
of each component that failed during the event; and

(ix) For events occurring during use of a packaging, the
quantities and chemical and physical form(s) of the package contents.

(3) An assessment of the safety consequences and implications of
the event. This assessment must include the availability of other
systems or components that could have performed the same function as
the components and systems that failed during the event.

(4) A description of any corrective actions planned as a result of
the event, including the means employed to repair any defects, and
actions taken to reduce the probability of similar events occurring
in the future.

(5) Reference to any previous similar events involving the same
packaging that are known to the licensee or certificate holder.

(6) The name and telephone number of a person within the
licensee's organization who is knowledgeable about the event and can
provide additional information.

(7) The extent of exposure of individuals to radiation or to
radioactive materials without identification of individuals by name.

(d) Report legibility. The reports submitted by licensees and/or
certificate holders under this section must be of sufficient quality
to permit reproduction and micrographic processing.

(a)(1) As specified in paragraphs (b), (c), and (d) of this
section, each licensee shall provide advance notification to the
governor of a State, or the governor’s designee, of the shipment of
licensed material, within or across the boundary of the State, before
the transport, or delivery to a carrier, for transport, of licensed
material outside the confines of the licensee’s plant or other
place of use or storage.

(2) As specified in paragraphs (b), (c), and (d) of this section,
after June 11, 2013, each licensee shall provide advance notification
to the Tribal official of participating Tribes referenced in
paragraph (c)(3)(iii) of this section, or the official’s designee,
of the shipment of licensed material, within or across the boundary
of the Tribe’s reservation, before the transport, or delivery to a
carrier, for transport, of licensed material outside the confines of
the licensee’s plant or other place of use or storage.

(b) Advance notification is also required under this section for
the shipment of licensed material, other than irradiated fuel,
meeting the following three conditions:

(1) The licensed material is required by this part to be in Type B
packaging for transportation;

(2) The licensed material is being transported to or across a
State boundary en route to a disposal facility or to a collection
point for transport to a disposal facility; and

(3) The quantity of licensed material in a single package exceeds
the least of the following:

(i) 3000 times the A1
value of the radionuclides as specified in appendix A, Table A–1
for special form radioactive material;

(ii) 3000 times the A2
value of the radionuclides as specified in appendix A, Table A–1
for normal form radioactive material; or

(iii) 1000 TBq (27,000 Ci).

(c) Procedures for submitting advance notification. (1)
The notification must be made in writing to:

(i) The office of each appropriate governor or governor’s
designee;

(ii) The office of each appropriate Tribal official or Tribal
official’s designee; and

(2) A notification delivered by mail must be postmarked at least 7
days before the beginning of the 7-day period during which departure
of the shipment is estimated to occur.

(3) A notification delivered by any other means than mail must
reach the office of the governor or of the governor’s designee or
the Tribal official or Tribal official’s designee at least 4 days
before the beginning of the 7-day period during which departure of
the shipment is estimated to occur.

(i) A list of the names and mailing addresses of the governors’
designees receiving advance notification of transportation of nuclear
waste was published in the Federal Register on June
30, 1995 (60 FR 34306).

(ii) The list of governor’s designees and Tribal official’s
designees of participating Tribes will be published annually in the
Federal Register on or about June 30th to reflect
any changes in information.

(iii) A list of the names and mailing addresses of the governors’
designees and Tribal officials’ designees of participating Tribes
is available on request from the Director, Division of
Intergovernmental Liaison and Rulemaking, Office of Federal and State
Materials and Environmental Management Programs, U.S. Nuclear
Regulatory Commission, Washington, DC 20555–0001.

(4) The licensee shall retain a copy of the notification as a
record for 3 years.

(d) Information to be furnished in advance notification of
shipment. Each advance notification of shipment of irradiated
reactor fuel or nuclear waste must contain the following information:

(1) The name, address, and telephone number of the shipper,
carrier, and receiver of the irradiated reactor fuel or nuclear waste
shipment;

(2) A description of the irradiated reactor fuel or nuclear waste
contained in the shipment, as specified in the regulations of DOT in
49 CFR 172.202 and 172.203(d);

(3) The point of origin of the shipment and the 7-day period
during which departure of the shipment is estimated to occur;

(4) The 7-day period during which arrival of the shipment at State
boundaries or Tribal reservation boundaries is estimated to occur;

(5) The destination of the shipment, and the 7-day period during
which arrival of the shipment is estimated to occur; and

(6) A point of contact, with a telephone number, for current
shipment information.

(e) Revision notice. A licensee who finds that schedule
information previously furnished to a governor or governor’s
designee or a Tribal official or Tribal official’s designee, in
accordance with this section, will not be met, shall telephone a
responsible individual in the office of the governor of the State or
of the governor’s designee or the Tribal official or the Tribal
official’s designee and inform that individual of the extent of the
delay beyond the schedule originally reported. The licensee shall
maintain a record of the name of the individual contacted for 3
years.

(f) Cancellation notice. (1) Each licensee who cancels an
irradiated reactor fuel or nuclear waste shipment for which advance
notification has been sent shall send a cancellation notice to the
governor of each State or to the governor’s designee previously
notified, each Tribal official or to the Tribal official’s designee
previously notified, and to the Director, Division of Security
Policy, Office of Nuclear Security and Incident Response.

(2) The licensee shall state in the notice that it is a
cancellation and identify the advance notification that is being
canceled. The licensee shall retain a copy of the notice as a record
for 3 years.

(iii) Any rule, regulation, or order issued pursuant to the
sections specified in paragraph (b)(1)(i) of this section; or

(iv) Any term , condition, or limitation of any license issued
under the sections specified in paragraph (b)(1)(i) of this section.

(2) For any violation for which a license may be revoked under
section 186 of the Atomic Energy Act of 1954, as amended.

§ 71.100 Criminal penalties.

(a) Section 223 of the Atomic Energy Act of 1954, as amended,
provides for criminal sanctions for willful violation of, attempted
violation of, or conspiracy to violate, any regulation issued under
sections 161b, 161i, or 161o of the Act. For purposes of section 223,
all the regulations in part 71 are issued under one or more of
sections 161b, 161i, or 161o, except for the sections listed in
paragraph (b) of this section.

Subpart H--Quality Assurance

Source: 69 FR 3797, Jan. 26, 2004, unless otherwise noted.

§ 71.101 Quality assurance requirements.

(a) Purpose. This subpart describes quality assurance
requirements applying to design, purchase, fabrication, handling,
shipping, storing, cleaning, assembly, inspection, testing,
operation, maintenance, repair, and modification of components of
packaging that are important to safety. As used in this subpart,
"quality assurance" comprises all those planned and
systematic actions necessary to provide adequate confidence that a
system or component will perform satisfactorily in service. Quality
assurance includes quality control, which comprises those quality
assurance actions related to control of the physical characteristics
and quality of the material or component to predetermined
requirements. The licensee, certificate holder, and applicant for a
CoC are responsible for the quality assurance requirements as they
apply to design, fabrication, testing, and modification of packaging.
Each licensee is responsible for the quality assurance provision
which applies to its use of a packaging for the shipment of licensed
material subject to this subpart.

(b) Establishment of program. Each licensee, certificate
holder, and applicant for a CoC shall establish, maintain, and
execute a quality assurance program satisfying each of the applicable
criteria of §§ 71.101 through 71.137 and satisfying any specific
provisions that are applicable to the licensee's activities including
procurement of packaging. The licensee, certificate holder, and
applicant for a CoC shall execute the applicable criteria in a graded
approach to an extent that is commensurate with the quality assurance
requirement's importance to safety.

(c) Approval of program. (1) Before the use of any
package for the shipment of licensed material subject to this
subpart, each licensee shall obtain Commission approval of its
quality assurance program. Using an appropriate method listed in §
71.1(a), each licensee shall file a description of its quality
assurance program, including a discussion of which requirements of
this subpart are applicable and how they will be satisfied, by
submitting the description to: ATTN: Document Control Desk, Director,
Division of Spent Fuel Storage and Transportation, Office of Nuclear
Material Safety and Safeguards.

(2) Before the fabrication, testing, or modification of any
package for the shipment of licensed material subject to this
subpart, each licensee, certificate holder, or applicant for a CoC
shall obtain Commission approval of its quality assurance program.
Each certificate holder or applicant for a CoC shall, in accordance
with § 71.1, file a description of its quality assurance program,
including a discussion of which requirements of this subpart are
applicable and how they will be satisfied.

(d) Existing package designs. The provisions of this
paragraph deal with packages that have been approved for use in
accordance with this part before January 1, 1979, and which have been
designed in accordance with the provisions of this part in effect at
the time of application for package approval. Those packages will be
accepted as having been designed in accordance with a quality
assurance program that satisfies the provisions of paragraph (b) of
this section.

(e) Existing packages. The provisions of this paragraph
deal with packages that have been approved for use in accordance with
this part before January 1, 1979, have been at least partially
fabricated before that date, and for which the fabrication is in
accordance with the provisions of this part in effect at the time of
application for approval of package design. These packages will be
accepted as having been fabricated and assembled in accordance with a
quality assurance program that satisfies the provisions of paragraph
(b) of this section.

(f) Previously approved programs. A Commission-approved
quality assurance program that satisfies the applicable criteria of
subpart H of this part, Appendix B of part 50 of this chapter, or
subpart G of part 72 of this chapter, and that is established,
maintained, and executed regarding transport packages, will be
accepted as satisfying the requirements of paragraph (b) of this
section. Before first use, the licensee, certificate holder, and
applicant for a CoC shall notify the NRC, in accordance with § 71.1,
of its intent to apply its previously approved subpart H, Appendix B,
or subpart G quality assurance program to transportation activities.
The licensee, certificate holder, and applicant for a CoC shall
identify the program by date of submittal to the Commission, Docket
Number, and date of Commission approval.

(g) Radiography containers. A program for transport
container inspection and maintenance limited to radiographic exposure
devices, source changers, or packages transporting these devices and
meeting the requirements of § 34.31(b) of this chapter or equivalent
Agreement State requirement, is deemed to satisfy the requirements of
§§ 71.17(b) and 71.101(b).

[75 FR 73945, Nov. 30, 2010]

§ 71.103 Quality assurance organization.

(a) The licensee,2
certificate holder, and applicant for a CoC shall be responsible for
the establishment and execution of the quality assurance program. The
licensee, certificate holder, and applicant for a CoC may delegate to
others, such as contractors, agents, or consultants, the work of
establishing and executing the quality assurance program, or any part
of the quality assurance program, but shall retain responsibility for
the program. These activities include performing the functions
associated with attaining quality objectives and the quality
assurance functions.

(b) The quality assurance functions are--

(1) Assuring that an appropriate quality assurance program is
established and effectively executed; and

(2) Verifying, by procedures such as checking, auditing, and
inspection, that activities affecting the functions that are
important to safety have been correctly performed.

(d) The persons and organizations performing quality assurance
functions shall report to a management level that assures that the
required authority and organizational freedom, including sufficient
independence from cost and schedule, when opposed to safety
considerations, are provided.

(e) Because of the many variables involved, such as the number of
personnel, the type of activity being performed, and the location or
locations where activities are performed, the organizational
structure for executing the quality assurance program may take
various forms, provided that the persons and organizations assigned
the quality assurance functions have the required authority and
organizational freedom.

(f) Irrespective of the organizational structure, the
individual(s) assigned the responsibility for assuring effective
execution of any portion of the quality assurance program, at any
location where activities subject to this section are being
performed, must have direct access to the levels of management
necessary to perform this function.

2
While the term "licensee" is used in these criteria, the
requirements are applicable to whatever design, fabrication,
assembly, and testing of the package is accomplished with respect to
a package before the time a package approval is issued.

§ 71.105 Quality assurance program.

(a) The licensee, certificate holder, and applicant for a CoC
shall establish, at the earliest practicable time consistent with the
schedule for accomplishing the activities, a quality assurance
program that complies with the requirements of §§ 71.101 through
71.137. The licensee, certificate holder, and applicant for a CoC
shall document the quality assurance program by written procedures or
instructions and shall carry out the program in accordance with those
procedures throughout the period during which the packaging is used.
The licensee, certificate holder, and applicant for a CoC shall
identify the material and components to be covered by the quality
assurance program, the major organizations participating in the
program, and the designated functions of these organizations.

(b) The licensee, certificate holder, and applicant for a CoC,
through its quality assurance program, shall provide control over
activities affecting the quality of the identified materials and
components to an extent consistent with their importance to safety,
and as necessary to assure conformance to the approved design of each
individual package used for the shipment of radioactive material. The
licensee, certificate holder, and applicant for a CoC shall assure
that activities affecting quality are accomplished under suitably
controlled conditions. Controlled conditions include the use of
appropriate equipment; suitable environmental conditions for
accomplishing the activity, such as adequate cleanliness; and
assurance that all prerequisites for the given activity have been
satisfied. The licensee, certificate holder, and applicant for a CoC
shall take into account the need for special controls, processes,
test equipment, tools, and skills to attain the required quality, and
the need for verification of quality by inspection and test.

(c) The licensee, certificate holder, and applicant for a CoC
shall base the requirements and procedures of its quality assurance
program on the following considerations concerning the complexity and
proposed use of the package and its components:

(1) The impact of malfunction or failure of the item to safety;

(2) The design and fabrication complexity or uniqueness of the
item;

(3) The need for special controls and surveillance over processes
and equipment;

(4) The degree to which functional compliance can be demonstrated
by inspection or test; and

(5) The quality history and degree of standardization of the item.

(d) The licensee, certificate holder, and applicant for a CoC
shall provide for indoctrination and training of personnel performing
activities affecting quality, as necessary to assure that suitable
proficiency is achieved and maintained. The licensee, certificate
holder, and applicant for a CoC shall review the status and adequacy
of the quality assurance program at established intervals. Management
of other organizations participating in the quality assurance program
shall review regularly the status and adequacy of that part of the
quality assurance program they are executing.

§ 71.107 Package design control.

(a) The licensee, certificate holder, and applicant for a CoC
shall establish measures to assure that applicable regulatory
requirements and the package design, as specified in the license or
CoC for those materials and components to which this section applies,
are correctly translated into specifications, drawings, procedures,
and instructions. These measures must include provisions to assure
that appropriate quality standards are specified and included in
design documents and that deviations from standards are controlled.
Measures must be established for the selection and review for
suitability of application of materials, parts, equipment, and
processes that are essential to the functions of the materials,
parts, and components of the packaging that are important to safety.

(b) The licensee, certificate holder, and applicant for a CoC
shall establish measures for the identification and control of design
interfaces and for coordination among participating design
organizations. These measures must include the establishment of
written procedures, among participating design organizations, for the
review, approval, release, distribution, and revision of documents
involving design interfaces. The design control measures must provide
for verifying or checking the adequacy of design, by methods such as
design reviews, alternate or simplified calculational methods, or by
a suitable testing program. For the verifying or checking process,
the licensee shall designate individuals or groups other than those
who were responsible for the original design, but who may be from the
same organization. Where a test program is used to verify the
adequacy of a specific design feature in lieu of other verifying or
checking processes, the licensee, certificate holder, and applicant
for a CoC shall include suitable qualification testing of a prototype
or sample unit under the most adverse design conditions. The
licensee, certificate holder, and applicant for a CoC shall apply
design control measures to the following:

(c) The licensee, certificate holder, and applicant for a CoC
shall subject design changes, including field changes, to design
control measures commensurate with those applied to the original
design. Changes in the conditions specified in the CoC require prior
NRC approval.

§ 71.109 Procurement document control.

The licensee, certificate holder, and applicant for a CoC shall
establish measures to assure that adequate quality is required in the
documents for procurement of material, equipment, and services,
whether purchased by the licensee, certificate holder, and applicant
for a CoC or by its contractors or subcontractors. To the extent
necessary, the licensee, certificate holder, and applicant for a CoC
shall require contractors or subcontractors to provide a quality
assurance program consistent with the applicable provisions of this
part.

§ 71.111 Instructions, procedures, and drawings.

The licensee, certificate holder, and applicant for a CoC shall
prescribe activities affecting quality by documented instructions,
procedures, or drawings of a type appropriate to the circumstances
and shall require that these instructions, procedures, and drawings
be followed. The instructions, procedures, and drawings must include
appropriate quantitative or qualitative acceptance criteria for
determining that important activities have been satisfactorily
accomplished.

§ 71.113 Document control.

The licensee, certificate holder, and applicant for a CoC shall
establish measures to control the issuance of documents such as
instructions, procedures, and drawings, including changes, that
prescribe all activities affecting quality. These measures must
assure that documents, including changes, are reviewed for adequacy,
approved for release by authorized personnel, and distributed and
used at the location where the prescribed activity is performed.

§ 71.115 Control of purchased material,
equipment, and services.

(a) The licensee, certificate holder, and applicant for a CoC
shall establish measures to assure that purchased material,
equipment, and services, whether purchased directly or through
contractors and subcontractors, conform to the procurement documents.
These measures must include provisions, as appropriate, for source
evaluation and selection, objective evidence of quality furnished by
the contractor or subcontractor, inspection at the contractor or
subcontractor source, and examination of products on delivery.

(b) The licensee, certificate holder, and applicant for a CoC
shall have available documentary evidence that material and equipment
conform to the procurement specifications before installation or use
of the material and equipment. The licensee, certificate holder, and
applicant for a CoC shall retain, or have available, this documentary
evidence for the life of the package to which it applies. The
licensee, certificate holder, and applicant for a CoC shall assure
that the evidence is sufficient to identify the specific requirements
met by the purchased material and equipment.

(c) The licensee, certificate holder, and applicant for a CoC
shall assess the effectiveness of the control of quality by
contractors and subcontractors at intervals consistent with the
importance, complexity, and quantity of the product or services.

§ 71.117 Identification and control of
materials, parts, and components.

The licensee, certificate holder, and applicant for a CoC shall
establish measures for the identification and control of materials,
parts, and components. These measures must assure that identification
of the item is maintained by heat number, part number, or other
appropriate means, either on the item or on records traceable to the
item, as required throughout fabrication, installation, and use of
the item. These identification and control measures must be designed
to prevent the use of incorrect or defective materials, parts, and
components.

§ 71.119 Control of special processes.

The licensee, certificate holder, and applicant for a CoC shall
establish measures to assure that special processes, including
welding, heat treating, and nondestructive testing are controlled and
accomplished by qualified personnel using qualified procedures in
accordance with applicable codes, standards, specifications,
criteria, and other special requirements.

§ 71.121 Internal inspection.

The licensee, certificate holder, and applicant for a CoC shall
establish and execute a program for inspection of activities
affecting quality by or for the organization performing the activity,
to verify conformance with the documented instructions, procedures,
and drawings for accomplishing the activity. The inspection must be
performed by individuals other than those who performed the activity
being inspected. Examination, measurements, or tests of material or
products processed must be performed for each work operation where
necessary to assure quality. If direct inspection of processed
material or products is not carried out, indirect control by
monitoring processing methods, equipment, and personnel must be
provided. Both inspection and process monitoring must be provided
when quality control is inadequate without both. If mandatory
inspection hold points, which require witnessing or inspecting by the
licensee's designated representative and beyond which work should not
proceed without the consent of its designated representative, are
required, the specific hold points must be indicated in appropriate
documents.

§ 71.123 Test control.

The licensee, certificate holder, and applicant for a CoC shall
establish a test program to assure that all testing required to
demonstrate that the packaging components will perform satisfactorily
in service is identified and performed in accordance with written
test procedures that incorporate the requirements of this part and
the requirements and acceptance limits contained in the package
approval. The test procedures must include provisions for assuring
that all prerequisites for the given test are met, that adequate test
instrumentation is available and used, and that the test is performed
under suitable environmental conditions. The licensee, certificate
holder, and applicant for a CoC shall document and evaluate the test
results to assure that test requirements have been satisfied.

§ 71.125 Control of measuring and test
equipment.

The licensee, certificate holder, and applicant for a CoC shall
establish measures to assure that tools, gauges, instruments, and
other measuring and testing devices used in activities affecting
quality are properly controlled, calibrated, and adjusted at
specified times to maintain accuracy within necessary limits.

§ 71.127 Handling, storage, and shipping
control.

The licensee, certificate holder, and applicant for a CoC shall
establish measures to control, in accordance with instructions, the
handling, storage, shipping, cleaning, and preservation of materials
and equipment to be used in packaging to prevent damage or
deterioration. When necessary for particular products, special
protective environments, such as inert gas atmosphere, and specific
moisture content and temperature levels must be specified and
provided.

§ 71.129 Inspection, test, and operating status.

(a) The licensee, certificate holder, and applicant for a CoC
shall establish measures to indicate, by the use of markings such as
stamps, tags, labels, routing cards, or other suitable means, the
status of inspections and tests performed upon individual items of
the packaging. These measures must provide for the identification of
items that have satisfactorily passed required inspections and tests,
where necessary to preclude inadvertent bypassing of the inspections
and tests.

(b) The licensee shall establish measures to identify the
operating status of components of the packaging, such as tagging
valves and switches, to prevent inadvertent operation.

§ 71.131 Nonconforming materials, parts, or
components.

The licensee, certificate holder, and applicant for a CoC shall
establish measures to control materials, parts, or components that do
not conform to the licensee's requirements to prevent their
inadvertent use or installation. These measures must include, as
appropriate, procedures for identification, documentation,
segregation, disposition, and notification to affected organizations.
Nonconforming items must be reviewed and accepted, rejected,
repaired, or reworked in accordance with documented procedures.

§ 71.133 Corrective action.

The licensee, certificate holder, and applicant for a CoC shall
establish measures to assure that conditions adverse to quality, such
as deficiencies, deviations, defective material and equipment, and
nonconformances, are promptly identified and corrected. In the case
of a significant condition adverse to quality, the measures must
assure that the cause of the condition is determined and corrective
action taken to preclude repetition. The identification of the
significant condition adverse to quality, the cause of the condition,
and the corrective action taken must be documented and reported to
appropriate levels of management.

§ 71.135 Quality assurance records.

The licensee, certificate holder, and applicant for a CoC shall
maintain sufficient written records to describe the activities
affecting quality. The records must include the instructions,
procedures, and drawings required by § 71.111 to prescribe quality
assurance activities and must include closely related specifications
such as required qualifications of personnel, procedures, and
equipment. The records must include the instructions or procedures
which establish a records retention program that is consistent with
applicable regulations and designates factors such as duration,
location, and assigned responsibility. The licensee, certificate
holder, and applicant for a CoC shall retain these records for 3
years beyond the date when the licensee, certificate holder, and
applicant for a CoC last engage in the activity for which the quality
assurance program was developed. If any portion of the written
procedures or instructions is superseded, the licensee, certificate
holder, and applicant for a CoC shall retain the superseded material
for 3 years after it is superseded.

§ 71.137 Audits.

The licensee, certificate holder, and applicant for a CoC shall
carry out a comprehensive system of planned and periodic audits to
verify compliance with all aspects of the quality assurance program
and to determine the effectiveness of the program. The audits must be
performed in accordance with written procedures or checklists by
appropriately trained personnel not having direct responsibilities in
the areas being audited. Audited results must be documented and
reviewed by management having responsibility in the area audited.
Followup action, including reaudit of deficient areas, must be taken
where indicated.

Appendix A to Part 71—Determination of A1
and A2

I. Values of A1
and A2 for
individual radionuclides, which are the bases for many activity
limits elsewhere in these regulations, are given in Table A-1. The
curie (Ci) values specified are obtained by converting from the
Terabecquerel (TBq) value. The Terabecquerel values are the
regulatory standard. The curie values are for information only and
are not intended to be the regulatory standard. Where values of A1
and A2 are
unlimited, it is for radiation control purposes only. For nuclear
criticality safety, some materials are subject to controls placed on
fissile material.

II. a. For individual radionuclides whose identities are known,
but which are not listed in Table A-1, the A1
and A2 values
contained in Table A-3 may be used. Otherwise, the licensee shall
obtain prior Commission approval of the A1
and A2 values
for radionuclides not listed in Table A-1, before shipping the
material.

b. For individual radionuclides whose identities are known, but
which are not listed in Table A-2, the exempt material activity
concentration and exempt consignment activity values contained in
Table A-3 may be used. Otherwise, the licensee shall obtain prior
Commission approval of the exempt material activity concentration and
exempt consignment activity values for radionuclides not listed in
Table A-2, before shipping the material.

c. The licensee shall submit requests for prior approval,
described under paragraphs II(a) and II(b) of this Appendix, to the
Commission, in accordance with § 71.1 of this part.

III. In the calculations of A1
and A2 for a
radionuclide not in Table A-1, a single radioactive decay chain, in
which radionuclides are present in their naturally occurring
proportions, and in which no daughter radionuclide has a half-life
either longer than 10 days, or longer than that of the parent
radionuclide, shall be considered as a single radionuclide, and the
activity to be taken into account, and the A1
orA2
value to be applied, shall be those corresponding to the parent
radionuclide of that chain. In the case of radioactive decay chains
in which any daughter radionuclide has a half-life either longer than
10 days, or greater than that of the parent radionuclide, the parent
and those daughter radionuclides shall be considered as mixtures of
different radionuclides.

IV. For mixtures of radionuclides whose identities and respective
activities are known, the following conditions apply:

a. For special form radioactive material, the maximum quantity
transported in a Type A package is as follows:

where B(i) is the activity of radionuclide i, and A1(i)
is the A1 value
for radionuclide I.

b. For normal form radioactive material, the maximum quantity
transported in a Type A package is as follows:

∑B(i)/A2(i)≤1

where B(i) is the activity of radionuclide i, and A2(i)
is the A2 value
for radionuclide i.

c. Alternatively, the A1
value for mixtures of special form material may be determined as
follows:

where f(i) is the fraction of activity for radionuclide I in the
mixture, and A1(i)
is the appropriate A1
value for radionuclide I.

d. Alternatively, the A2
value for mixtures of normal form material may be determined as
follows:

where f(i) is the fraction of activity for radionuclide I in the
mixture, and A2(i)
is the appropriate A2
value for radionuclide I.

e. The exempt activity concentration for mixtures of nuclides may
be determined as follows:

where f(i) is the fraction of activity concentration of
radionuclide I in the mixture, and [A] is the activity concentration
for exempt material containing radionuclide I.

f. The activity limit for an exempt consignment for mixtures of
radionuclides may be determined as follows:

where f(i) is the fraction of activity of radionuclide I in the
mixture, and A is the activity limit for exempt consignments for
radionuclide I.

V. When the identity of each radionuclide is known, but the
individual activities of some of the radionuclides are not known, the
radionuclides may be grouped, and the lowest A1
or A2 value, as
appropriate, for the radionuclides in each group may be used in
applying the formulas in paragraph IV. Groups may be based on the
total alpha activity and the total beta/gamma activity when these are
known, using the lowest A1
or A2 values
for the alpha emitters and beta/gamma emitters.

a A1 and/or A2
values include contributions from daughter nuclides with half-lives
less than 10 days.
b The values of A1
and A2 in
Curies (Ci) are approximate and for information only; the regulatory
standard units are Terabecquerels (TBq) (see Appendix A to Part
71—Determination of A1
and A2, Section
I).c The
quantity may be determined from a measurement of the rate of decay or
a measurement of the radiation level at a prescribed distance from
the source.d
These values apply only to compounds of uranium that take the
chemical form of UF6,
UO2F2
and UO2(NO3)2
in both normal and accident conditions of transport.e
These values apply only to compounds of uranium that take the
chemical form of UO3,
UF4, UCl4
and hexavalent compounds in both normal and accident conditions of
transport.f
These values apply to all compounds of uranium other than those
specified in notes (d) and (e) of this table.g
These values apply to unirradiated uranium only.h
A1 = 0.1 TBq
(2.7 Ci) and A2
= 0.001 TBq (0.027 Ci) for Cf-252 for domestic use.i
A2 = 0.74 TBq
(20 Ci) for Mo-99 for domestic use.

c [Reserved]d
These values apply only to compounds of uranium that take the
chemical form of UF6,
UO2F2
and UO2(NO3)2
in both normal and accident conditions of transport.e
These values apply only to compounds of uranium that take the
chemical form of UO3,
UF4, UCl4
and hexavalent compounds in both normal and accident conditions of
transport.f
These values apply to all compounds of uranium other than those
specified in notes (d) and (e) of this table.g
These values apply to unirradiated uranium only.